ITER tokamak device

International Nuclear Information System (INIS)

Doggett, J.; Salpietro, E.; Shatalov, G.

1991-01-01

The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

Assembly and gap management strategy for the ITER NBI vessel passive magnetic shield

Energy Technology Data Exchange (ETDEWEB)

Ríos, Luis, E-mail: luis.rios@ciemat.es [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Ahedo, Begoña; Alonso, Javier; Barrera, Germán; Cabrera, Santiago; Rincón, Esther; Ramos, Francisco [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); El-Ouazzani, Anass; Graceffa, Joseph; Urbani, Marc; Shah, Darshan [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3 – 07/08, 08019 Barcelona (Spain)

2015-10-15

The neutral beam system for ITER consists of two heating and current drive neutral ion beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. The HNB Passive Magnetic Shield (PMS) works in conjunction with the active compensation/correction coils to limit the magnetic field inside the Beam Line Vessel (BLV), Beam Source Vessel (BSV), High Voltage Bushing (HVB) and Transmission Line (TL) elbow to acceptable levels that do not interfere with the operation of the HNB components. This paper describes the current design of the PMS, having had only minor modifications since the preliminary design review (PDR) held in IO in April 2013, and the assembly strategy for the vessel PMS.

Design of the ITER high-frequency magnetic diagnostic coils

International Nuclear Information System (INIS)

Toussaint, M.; Testa, D.; Baluc, N.; Chavan, R.; Fournier, Y.; Lister, J.B.; Maeder, T.; Marmillod, P.; Sanchez, F.; Stoeck, M.

2011-01-01

This paper is an overview of work carried out on the design of the ITER high-frequency magnetic diagnostic coil (HF sensor). In the first part, the ITER requirements for the HF sensor are presented. In the second part, the ITER reference design of the HF sensor has been assessed and showed some potential weaknesses, which led us to the conclusion that alternative designs could usefully be examined. Several options have been explored, and are presented in the third part: (a) direct laser cutting a metallic tube, (b) stacking of plane windings manufactured from a tungsten plate by electrical discharge machining, (c) coil using the conventional spring manufacture. In the fourth part, sensors using the low temperature co-fired ceramic technology (LTCC) are presented: (d) monolithic 1D magnetic flux sensors based on LTCC technology, and (e) monolithic 3D magnetic flux sensors based on the same LTCC technology. The solution which showed the best results is the monolithic 3D magnetic flux sensor based on LTCC.

Design and development of ITER high-frequency magnetic sensor

NARCIS (Netherlands)

Ma, Y.; Vayakis, G.; Begrambekov, L. B.; Cooper, J.J.; Duran, I.; Hirsch, M.; Laqua, H.P.; Moreau, Ph.; Oosterbeek, J.W.; Spuig, P.; Stange, T.; Walsh, M.

2016-01-01

High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in

Status of the ITER magnets

International Nuclear Information System (INIS)

Mitchell, N.; Bauer, P.; Bessette, D.; Devred, A.; Gallix, R.; Jong, C.; Knaster, J.; Libeyre, P.; Lim, B.; Sahu, A.; Simon, F.

2009-01-01

The first 2 years of the ITER IO has seen substantial progress towards the construction of the magnets, in three main areas. Firstly, the design has been developed under the conflicting constraints to minimise construction costs and to maximise plasma physics performance. Building construction momentum while updating the design to take account of new physics assessments of the coil requirements has been challenging. Secondly, with a stabilising design, it has been possible for the Domestic Agencies to launch the first industrial procurement contracts. And thirdly, critical R and D to confirm the performance of the Nb3Sn cable in conduit design is proceeding successfully. The design consolidation has been accompanied by design reviews involving the international community. The reviews conducted by magnet experts have enabled a consensus to be built on choosing between some of the design options in the original ITER basic design in 2001. The major design decisions were to maintain the circular Nb 3 Sn conductor embedded in radial plates for the toroidal field (TF) coils and to maintain NbTi-based conductors for the PF coils. Cold testing, at low current, is also being introduced for quality control purposes for all coils.

Engineering, Manufacture and Preliminary Testing of the ITER Toroidal Field (TF) Magnet Helium Cold Circulator

Science.gov (United States)

Rista, P. E. C.; Shull, J.; Sargent, S.

2015-12-01

The ITER cryodistribution system provides the supercritical Helium (SHe) forced flow cooling to the magnet system using cold circulators. The cold circulators are located in each of five separate auxiliary cold boxes planned for use in the facility. Barber-Nichols Inc. has been awarded a contract from ITER-India for engineering, manufacture and testing of the Toroidal Field (TF) Magnet Helium Cold Circulator. The cold circulator will be extensively tested at Barber-Nichols’ facility prior to delivery for qualification testing at the Japan Atomic Energy Agency's (JAEA) test facility at Naka, Japan. The TF Cold Circulator integrates features and technical requirements which Barber-Nichols has utilized when supplying helium cold circulators worldwide over a period of 35 years. Features include a vacuum-jacketed hermetically sealed design with a very low helium leak rate, a heat shield for use with both nitrogen & helium cold sources, a broad operating range with a guaranteed isentropic efficiency over 70%, and impeller design features for high efficiency. The cold circulator will be designed to meet MTBM of 17,500 hours and MTBF of 36,000 hours. Vibration and speed monitoring are integrated into a compact package on the rotating assembly with operation and health monitoring in a multi-drop PROFIBUS communication environment using an electrical cabinet with critical features and full local and network PLC interface and control. For the testing in Japan and eventual installation in Europe, the cold circulator must be certified to the Japanese High Pressure Gas Safety Act (JHPGSA) and CE marked in compliance with the European Pressure Equipment Directive (PED) including Essential Safety Requirements (ESR). The test methodology utilized at Barber-Nichols’ facility and the resulting test data, validating the high efficiency of the TF Cold Circulator across a broad operating range, are important features of this paper.

Engineering, Manufacture and Preliminary Testing of the ITER Toroidal Field (TF) Magnet Helium Cold Circulator

International Nuclear Information System (INIS)

C Rista, P E; Shull, J; Sargent, S

2015-01-01

The ITER cryodistribution system provides the supercritical Helium (SHe) forced flow cooling to the magnet system using cold circulators. The cold circulators are located in each of five separate auxiliary cold boxes planned for use in the facility. Barber-Nichols Inc. has been awarded a contract from ITER-India for engineering, manufacture and testing of the Toroidal Field (TF) Magnet Helium Cold Circulator. The cold circulator will be extensively tested at Barber-Nichols’ facility prior to delivery for qualification testing at the Japan Atomic Energy Agency's (JAEA) test facility at Naka, Japan. The TF Cold Circulator integrates features and technical requirements which Barber-Nichols has utilized when supplying helium cold circulators worldwide over a period of 35 years. Features include a vacuum-jacketed hermetically sealed design with a very low helium leak rate, a heat shield for use with both nitrogen and helium cold sources, a broad operating range with a guaranteed isentropic efficiency over 70%, and impeller design features for high efficiency. The cold circulator will be designed to meet MTBM of 17,500 hours and MTBF of 36,000 hours. Vibration and speed monitoring are integrated into a compact package on the rotating assembly with operation and health monitoring in a multi-drop PROFIBUS communication environment using an electrical cabinet with critical features and full local and network PLC interface and control. For the testing in Japan and eventual installation in Europe, the cold circulator must be certified to the Japanese High Pressure Gas Safety Act (JHPGSA) and CE marked in compliance with the European Pressure Equipment Directive (PED) including Essential Safety Requirements (ESR). The test methodology utilized at Barber-Nichols’ facility and the resulting test data, validating the high efficiency of the TF Cold Circulator across a broad operating range, are important features of this paper. (paper)

Engineering challenges and solutions for the ITER magnetic diagnostics flux loops

International Nuclear Information System (INIS)

Clough, M.; Casal, N.; Suarez Diaz, A.; Vayakis, G.; Walsh, M.

2014-01-01

The Magnetic Diagnostics Flux Loops (MDFL) are a key diagnostic for the ITER tokamak, providing important information about the shape of the plasma boundary, instabilities and magnetic error fields. In total, 237 flux loops will be installed on ITER, on the inside and outside walls of the Vacuum Vessel, and will range in area from 1 m 2 to 250 m 2 . This paper describes the detailed engineering design of the MDFL, explaining the solutions developed to maintain measurement accuracy within their difficult operating environment and other requirements: ultra-high vacuum conditions, strong magnetic fields, high gamma and neutron radiation doses, challenging installation, very high reliability and no maintenance during the 20 year machine lifetime. In addition, the paper discusses testing work undertaken to validate the design and outlines the remaining tasks to be performed. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. (authors)

Quality control in the design, fabrication and operation of the ITER magnets

International Nuclear Information System (INIS)

Mitchell, N.

2006-01-01

The ITER magnets are a complex system involving interfaces between many advanced technologies (superconductors, forging/welding/machining of massive structures, cryogenics, composites and moulding, high voltage electrical), yet at the same time form part of the ITER 'basic machine' which is required to operate at the design parameters, broadly failure free, for the design life of the tokamak. This imposes special quality control problems for the ITER project integration by the ITER International Team (IT) through the design, fabrication and operation. The magnets are not a test bed for new technology but in spite of this must use it, successfully. There is little previous experience of such a system but full functionality is required from the start, with limited opportunity for adjustment. And, finally, costs and schedule must be contained. The procurement strategy for the machine, with magnet components being supplied 'in kind', requires particular attention to the specifications, scheduling and quality control (QC). Special issues here are the testing requirements on magnet components, especially before final installation but also at critical intermediate stages. Unnecessary or ineffective quality control procedures cause delay and high costs, and divert attention from critical items. The main points of the magnet QC programme are summarised, including the use of codes and standards, qualification, manufacturing quality assurance, commissioning and in-service inspection

R and D on ITER in-vessel magnetic sensors

Energy Technology Data Exchange (ETDEWEB)

Peruzzo, Simone [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Arshad, Shakeib [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Brombin, Matteo, E-mail: matteo.brombin@igi.cnr.it [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Chitarin, Giuseppe; Gonzalez, Winder; Grando, Luca [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Portales, Mickael [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Rizzolo, Andrea [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Vayakis, George [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Vermeeren, Ludo [SCK-CEN, Institute of Advanced Nuclear Systems, Boeretang 200, B-2400 Mol (Belgium)

2013-10-15

Highlights: ► Magnetic sensors based on LTCC technology should be able to meet ITER requirements. ► Outgassing, TIEMF and RIEMF tests are planned on new prototypes. ► The design of a proper sensor support and connection system is in progress. -- Abstract: This paper summarizes the progress of R and D activities related to the design of in-vessel magnetic sensors for measurement of the equilibrium field in ITER and the associated mechanical support and connection system. The background to the design activities, developed in the last few years in the framework of several collaborations, is provided in the introduction. The paper focuses on the experimental results obtained from a set of sensor prototypes, recently manufactured and tested, and on numerical simulations performed in order to verify the compliance of the sensors with ITER requirements. The paper finally illustrates the status of the engineering analyses performed to progress the design of the support and connection system, conceived to be replaceable by remote handling. In conclusion, specific issues which require further developments to achieve the final design are outlined.

R and D on ITER in-vessel magnetic sensors

International Nuclear Information System (INIS)

Peruzzo, Simone; Arshad, Shakeib; Brombin, Matteo; Chitarin, Giuseppe; Gonzalez, Winder; Grando, Luca; Portales, Mickael; Rizzolo, Andrea; Vayakis, George; Vermeeren, Ludo

2013-01-01

Highlights: ► Magnetic sensors based on LTCC technology should be able to meet ITER requirements. ► Outgassing, TIEMF and RIEMF tests are planned on new prototypes. ► The design of a proper sensor support and connection system is in progress. -- Abstract: This paper summarizes the progress of R and D activities related to the design of in-vessel magnetic sensors for measurement of the equilibrium field in ITER and the associated mechanical support and connection system. The background to the design activities, developed in the last few years in the framework of several collaborations, is provided in the introduction. The paper focuses on the experimental results obtained from a set of sensor prototypes, recently manufactured and tested, and on numerical simulations performed in order to verify the compliance of the sensors with ITER requirements. The paper finally illustrates the status of the engineering analyses performed to progress the design of the support and connection system, conceived to be replaceable by remote handling. In conclusion, specific issues which require further developments to achieve the final design are outlined

Design and development of ITER high-frequency magnetic sensor

Energy Technology Data Exchange (ETDEWEB)

Ma, Y., E-mail: Yunxing.Ma@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Fircroft Engineering, Lingley House, 120 Birchwood Point, Birchwood Boulevard, Warrington, WA3 7QH (United Kingdom); Vayakis, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Begrambekov, L.B. [National Research Nuclear University (MEPhI), 115409, Moscow, Kashirskoe shosse 31 (Russian Federation); Cooper, J.-J. [Culham Centre for Fusion Energy (CCFE), Abingdon, Oxfordshire OX14 3DB (United Kingdom); Duran, I. [IPP Prague, Za Slovankou 1782/3, 182 00 Prague 8 (Czech Republic); Hirsch, M.; Laqua, H.P. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Moreau, Ph. [CEA Cadarache, 13108 Saint Paul lez Durance Cedex (France); Oosterbeek, J.W. [Eindhoven University of Technology (TU/e), PO Box 513, 5600 MB Eindhoven (Netherlands); Spuig, P. [CEA Cadarache, 13108 Saint Paul lez Durance Cedex (France); Stange, T. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Walsh, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

2016-11-15

Highlights: • ITER high-frequency magnetic sensor system has been designed. • Prototypes have been successfully manufactured. • Manufactured prototypes have been tested in various labs. • Test results experimentally validated the design. - Abstract: High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in a harsh environment with considerable neutron/nuclear radiation and high thermal load. Essential components of the HF sensor system, including inductive coil, electron cyclotron heating (ECH) shield, electrical cabling and termination load, have been designed to meet ITER measurement requirements. System performance (e.g. frequency response, thermal conduction) has been assessed. A prototyping campaign was initiated to demonstrate the manufacturability of the designed components. Prototypes have been produced according to the specifications. A series of lab tests have been performed to examine assembly issues and validate electrical and thermo-mechanical aspects of the design. In-situ microwave radiation test has been conducted in the MISTRAL test facility at IPP-Greifswald to experimentally examine the microwave shielding efficiency and structural integrity of the ECH shield. Low-power microwave attenuation measurement and scanning electron microscopic inspection were conducted to probe and examine the quality of the metal coating on the ECH shield.

Reactor structure and superconducting magnet system of ITER

International Nuclear Information System (INIS)

Tada, Eisuke; Yoshida, Kiyoshi; Shibanuma, Kiyoshi; Okuno, Kiyoshi; Tsuji, Hiroshi; Shimamoto, Susumu

1993-01-01

Fusion Experimental Reactors are one of the major steps toward realization of the fusion energy and the key objective are to demonstrate the scientific and technological feasibility prior to the Demo Fusion Reactor. ITER (International Thermonuclear Experimental Reactor) is one of experimental reactors and the conceptual design has been completed by the united efforts of USA, USSR, EC and Japan. In parallel with the conceptual design, key technology development in various areas has being conducted. This paper describes the overall design concepts and the latest technological achievements of the ITER reactor structure and superconducting magnet system. (author)

Review of design principles for ITER VV remote inspection in magnetic field

International Nuclear Information System (INIS)

Izard, Jean-Baptiste; Perrot, Yann; Friconneau, Jean-Pierre

2009-01-01

Because ITER magnet system has a limited number of mechanical and thermal stress cycles, shut down number of the toroidal field is limited during lifetime of ITER. Any inspection device able to withstand the toroidal field between two plasma shots will enhance the inspection frequency capacity of ITER during operation phase. In addition to the high magnetic field the system should also cope with high temperature, ultra-high vacuum and high radiation, in order to keep the reactor availability high. Radiation, ultra-high vacuum and temperature constraints already addressed by on going R and D activities within Europe-considering the required level of radiation is to date the highest encountered in remote handling, and that facing all these constraints at once is an additional issue to overcome. Whereas, operating remote handling systems in high magnetic field is quite new field of investigation. This paper aims to be a guideline for future designers to help them choose among options the adequate solution for an ITER relevant inspection device. It provides the designer an objective view of the different effects that stem from technical choices and help them deciding whether a technology is relevant or not depending on the task's requirements. We have selected a set of technologies and products available for structural design, actuation, sensing and data transmission in order to design inspection remote handling equipment for ITER in the given constraints. These different solutions are commented with specific considerations and directions to have them fit in the specifications. Different design strategies to cope with magnetic field are then discussed, which imply either insensitive design or using the magnetic field as a potential energy source and as a positioning help. This analysis is the first result of one of the projects in the PREFIT partnership, part of the European Fusion Training Scheme.

The international thermonuclear reactor (ITER)

International Nuclear Information System (INIS)

Fowler, T.K.; Henning, C.D.

1987-01-01

Four governmental groups, representing Europe, Japan, USSR and U.S. met in March 1987 to consider a new international design of a magnetic fusion device for the 1990's. An interim group was appointed. The author gives a brief synopsis of what might be thought of as a draft charter. The starting point is the objective of the ITER device, which is summarized as demonstrating both scientific and technical feasibility of fusion. The paper presents an update on the current thinking and technical aspects for the International Thermonuclear Experimental Reactor (ITER). This covers not only what is happening in the U.S. but also some reports of preliminary thinking of the last technical work that occurred in Vienna

Design and development of ITER high-frequency magnetic sensor

Czech Academy of Sciences Publication Activity Database

Ma, Y.; Vayakis, G.; Begrambekov, L.B.; Cooper, J.-J.; Ďuran, Ivan; Hirsch, M.; Laqua, H.P.; Moreau, P.; Oosterbeek, J.W.; Spuig, P.; Stange, T.; Walsh, M.

2016-01-01

Roč. 112, November (2016), s. 594-612 ISSN 0920-3796 Institutional support: RVO:61389021 Keywords : ITER * High-frequency * Magnetic diagnostics * ECHa Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016

Towards a preliminary design of the ITER plasma control system architecture

International Nuclear Information System (INIS)

Treutterer, W.; Rapson, C.J.; Raupp, G.; Snipes, J.; Vries, P. de; Winter, A.; Humphreys, D.A.; Walker, M.; Tommasi, G. de; Cinque, M.; Bremond, S.; Moreau, P.; Nouailletas, R.; Felton, R.

2017-01-01

Highlights: • ITER control requirements and use scenarios for initial plasma operation have been analysed. • Basic choices from conceptual design could be confirmed. • Architectural design considers dynamic structure changes. • All PCS components are integrated in an exception handling hierarchy. - Abstract: Design of the ITER plasma control system is proceeding towards its next – preliminary design – stage. During the conceptual design in 2013 an overall assessment of high-level control tasks and their relationships has been conducted. The goal of the preliminary design is to show, that a reasonable implementation of the proposed concepts exists which fulfills the high-level requirements and is suitable for realistic use cases. This verification is conducted with focus on the concrete use cases of early operation and first plasma, since these phases are mandatory for ITER startup. In particular, detailed control requirements and functions for commissioning and first plasma operation including breakdown, burn-through and ramp-up in L-mode, as well as for planned or exceptional shutdown are identified. Control functions related to those operational phases and the underlying control system architecture are modeled. The goal is to check whether the flexibility of the conceptual architectural approach is adequate also in consideration of the more elaborate definitions for control functions and their interactions. In addition, architecture shall already be prepared for extension to H-mode operation and burn-control, even if the related control functions are only roughly defined at the moment. As a consequence, the architectural design is amended where necessary and converted into base components and infrastructure services allowing to deploy control and exception handling algorithms for the concrete first-plasma operation.

Towards a preliminary design of the ITER plasma control system architecture

Energy Technology Data Exchange (ETDEWEB)

Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Rapson, C.J.; Raupp, G. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Snipes, J.; Vries, P. de; Winter, A. [ITER Organization, Route de Vinon sur Verdon, 13067 St Paul Lez Durance (France); Humphreys, D.A.; Walker, M. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Tommasi, G. de; Cinque, M. [CREATE/Università di Napoli Federico II, Napoli (Italy); Bremond, S.; Moreau, P.; Nouailletas, R. [Association CEA pour la Fusion Contrôlée, CEA Cadarache, 13108 St Paul les Durance (France); Felton, R. [CCFE Fusion Association, Culham Centre for Fusion Energy, Culham Science Centre, Oxfordshire, OX14 3DB (United Kingdom)

2017-02-15

Highlights: • ITER control requirements and use scenarios for initial plasma operation have been analysed. • Basic choices from conceptual design could be confirmed. • Architectural design considers dynamic structure changes. • All PCS components are integrated in an exception handling hierarchy. - Abstract: Design of the ITER plasma control system is proceeding towards its next – preliminary design – stage. During the conceptual design in 2013 an overall assessment of high-level control tasks and their relationships has been conducted. The goal of the preliminary design is to show, that a reasonable implementation of the proposed concepts exists which fulfills the high-level requirements and is suitable for realistic use cases. This verification is conducted with focus on the concrete use cases of early operation and first plasma, since these phases are mandatory for ITER startup. In particular, detailed control requirements and functions for commissioning and first plasma operation including breakdown, burn-through and ramp-up in L-mode, as well as for planned or exceptional shutdown are identified. Control functions related to those operational phases and the underlying control system architecture are modeled. The goal is to check whether the flexibility of the conceptual architectural approach is adequate also in consideration of the more elaborate definitions for control functions and their interactions. In addition, architecture shall already be prepared for extension to H-mode operation and burn-control, even if the related control functions are only roughly defined at the moment. As a consequence, the architectural design is amended where necessary and converted into base components and infrastructure services allowing to deploy control and exception handling algorithms for the concrete first-plasma operation.

Conceptual design of SC magnet system for ITER, (5)

International Nuclear Information System (INIS)

Nakajima, Hideo; Nishi, Masataka; Yoshida, Kiyoshi; Tsuji, Hiroshi; Egusa, Shigenori; Seguchi, Tadao; Hagiwara, Miyuki; Kirk, M.A.; Birtcher, R.C.

1991-08-01

Japan Atomic Energy Research Institute (JAERI) has been developing a superconducting magnet system for a fusion reactor. One of the key items in developing the superconducting magnets is material development and evaluation. The data of superconducting materials, structural alloys, and non-metallic materials are generated to establish a material data base at JAERI. This report is prepared to provide available data generated by JAERI to designers of superconducting magnets throughout the world. The following review papers written for the International Thermonuclear Experimental Reactor (ITER) report on conceptual design of magnet system are combined here. I. Superconducting Material Data II. Mechanical Properties of the Japanese Cryogenic Steels (JCS) at Cryogenic Temperature III. Review of Radiation Degradation Studies at JAERI on Composite Organic Insulators Used in Fusion Magnets (author)

Quench detection and behaviour in case of quench in the ITER magnet systems

International Nuclear Information System (INIS)

Coatanea-Gouachet, M.

2012-02-01

The quench of one of the ITER magnet system is an irreversible transition from superconducting to normal resistive state, of a conductor. This normal zone propagates along the cable in conduit conductor dissipating a large power. The detection has to be fast enough to dump out the magnetic energy and avoid irreversible damage of the systems. The primary quench detection in ITER is based on voltage detection, which is the most rapid detection. The very magnetically disturbed environment during the plasma scenario makes the voltage detection particularly difficult, inducing large inductive components in the coils and voltage compensations have to be designed to discriminate the resistive voltage associated with the quench. A conceptual design of the quench detection based on voltage measurements is proposed for the three majors magnet systems of ITER. For this, a clear methodology was developed. It includes the classical hot spot criterion, the quench propagation study using the commercial code Gandalf and the careful estimation of the inductive disturbances by developing the TrapsAV code. Specific solutions have been proposed for the compensation in the three ITER magnet systems and for the quench detection parameters, which are the voltage threshold (in the range of 0.1 V - 0.55 V) and the holding time (in the range of 1-1.4 s). The selected values, in particular the holding time, are sufficiently high to ensure the reliability of the system and avoid fast safety discharges not induced by a quench, which is a classical problem. (author)

Conceptual design of SC magnet system for ITER, (2)

International Nuclear Information System (INIS)

Koizumi, Koichi; Hasegawa, Mitsuru; Yoshida, Kiyoshi

1991-08-01

The International Thermonuclear Experimental Reactor (ITER) is an experimental tokamak machine testing the basic plasma performance and technologies required for future tokamak reactor. The design proposals for the Superconducting (SC) Magnet System from Japan were summarized by the Fusion Experimental Reactor (FER) Design Team and the Superconducting Magnet Laboratory of the Japan Atomic Energy Research Institute (JAERI). This report is one of the series reports on 'Conceptual design of superconducting magnet system for ITER', and describes the major results of the stress analysis regarding the Toroidal Field (TF) coil, the Center Solenoid (CS) coil and the Equilibrium Field (EF) coil and their support structures. Among the design issues, the mechanical design of the coil system was one of the most critical items, not only because of the huge electromagnetic loads due to large size and high magnetic field, but also because of the demand of high reliability under neutron irradiation. In order to satisfy both the coil performance and the mechanical reliability, different types of conductors were employed for each coils. The mechanical behaviors and the safety margin of each coil were analyzed by using finite element method (FEM) of MSC/NASTRAN. The procedure to obtain the equivalent winding stiffness employed for the each FEM analysis is also described in this report. The details on the coil specifications, conductor design and mechanical design for each coils are described in other report of the series reports. (J.P.N.)

Magnetic fusion and project ITER

International Nuclear Information System (INIS)

Park, H.K.

1992-01-01

It has already been demonstrated that our economics and international relationship are impacted by an energy crisis. For the continuing prosperity of the human race, a new and viable energy source must be developed within the next century. It is evident that the cost will be high and will require a long term commitment to achieve this goal due to a high degree of technological and scientific knowledge. Energy from the controlled nuclear fusion is a safe, competitive, and environmentally attractive but has not yet been completely conquered. Magnetic fusion is one of the most difficult technological challenges. In modem magnetic fusion devices, temperatures that are significantly higher than the temperatures of the sun have been achieved routinely and the successful generation of tens of million watts as a result of scientific break-even is expected from the deuterium and tritium experiment within the next few years. For the practical future fusion reactor, we need to develop reactor relevant materials and technologies. The international project called ''International Thermonuclear Experimental Reactor (ITER)'' will fulfill this need and the success of this project will provide the most attractive long-term energy source for mankind

Test of piezo-ceramic motor technology in ITER relevant high magnetic fields

Energy Technology Data Exchange (ETDEWEB)

Monti, Chiara, E-mail: chiara.monti@enea.it [Associazione EURATOM-ENEA sulla Fusione, via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Besi Vetrella, Ugo; Mugnaini, Giampiero; Neri, Carlo; Rossi, Paolo; Viola, Rosario [Associazione EURATOM-ENEA sulla Fusione, via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Dubus, Gregory; Damiani, Carlo [Fusion for Energy, c/ Josep Pla, 2 Torres Diagonal Litoral, 08019 Barcelona (Spain)

2014-10-15

In the framework of a Fusion for Energy (F4E) grant, a test campaign started in 2012 in order to assess the performance of the in-vessel viewing system (IVVS) probe concept and to verify its compatibility when exposed to ITER typical working conditions. ENEA laboratories went through with several tests simulating high magnetic fields, high temperature, high vacuum, gamma radiation and neutron radiation. A customized motor has been adopted to study the performances of ultrasonic piezo motors technology in high magnetic field conditions. This paper reports on the testing activity performed on the motor in a multi Tesla magnetic field. The job was carried out in a test facility of ENEA laboratories able to achieve 14 T. A maximum field of 10 T, fully compliant with ITER requirements (8 T), was applied. A specific mechanical assembly has been designed and manufactured to hold the motor in the region with high homogeneity of the field. Results obtained so far indicate that the motor is compatible with high magnetic fields, and are presented in the paper.

Preliminary Design Progress of the HCCR TBM for ITER testing

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Won; Park, Sung Dae; Kim, Dong Jun; Jin, Hyung Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Korea has designed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) including the TBM-shield, which is called the TBM-set, to be tested in ITER, a Nuclear Facility INB-174. Through the conceptual design review (CDR), its design integrity was successfully demonstrated at the conceptual design level at various loads. After CD approval, preliminary design (PD) was started and the progress is introduced in the present study. After PD review and approval, final design and then fabrication will be started. The main purpose of PD is to design the TBM-set according to the fabrication aspect and more detailed design for interfaces with ITER machine, such as installed TBM port plug and frame. With these considering, PD of TBM-set was started. PD for HCCR TBM has been performed (so far v0.24) from the CD model. FW, BZ, SW, TES/NAS, BM, and connecting support design were performed through the analyses, if necessary. The manufacturability was the main concern for PD model development. Thermal hydraulic analysis will be performed to evaluate the temperature and pressure drop in TBM-set. The structural integrity of TBM-set will be confirmed with combined various loads condition.

The integrated design of the ITER magnets and their auxiliary systems

International Nuclear Information System (INIS)

Huget, M.

1999-01-01

The magnet system design for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration to meet performance and operation requirements, including reliability and maintainability, in a cost effective manner. This paper identifies the requirements of long inductive burn time, large number of tokamak pulses, operational flexibility for the poloidal field (PF) system, magnet reliability and the cost constraints as the main design drivers. Key features of the magnet system which stem from these design drivers are described, together with interfaces and integration aspects of certain auxiliary systems. (author)

The integrated design of the ITER magnets and their auxiliary systems

International Nuclear Information System (INIS)

Huguet, M.

2001-01-01

The magnet system design for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration to meet performance and operation requirements, including reliability and maintainability, in a cost effective manner. This paper identifies the requirements of long inductive burn time, large number of tokamak pulses, operational flexibility for the poloidal field (PF) system, magnet reliability and the cost constraints as the main design drivers. Key features of the magnet system which stem from these design drivers are described, together with interfaces and integration aspects of certain auxiliary systems. (author)

Finding the magnetic size distribution of magnetic nanoparticles from magnetization measurements via the iterative Kaczmarz algorithm

Energy Technology Data Exchange (ETDEWEB)

Schmidt, Daniel, E-mail: frank.wiekhorst@ptb.de; Eberbeck, Dietmar; Steinhoff, Uwe; Wiekhorst, Frank

2017-06-01

The characterization of the size distribution of magnetic nanoparticles is an important step for the evaluation of their suitability for many different applications like magnetic hyperthermia, drug targeting or Magnetic Particle Imaging. We present a new method based on the iterative Kaczmarz algorithm that enables the reconstruction of the size distribution from magnetization measurements without a priori knowledge of the distribution form. We show in simulations that the method is capable of very exact reconstructions of a given size distribution and, in that, is highly robust to noise contamination. Moreover, we applied the method on the well characterized FeraSpin™ series and obtained results that were in accordance with literature and boundary conditions based on their synthesis via separation of the original suspension FeraSpin R. It is therefore concluded that this method is a powerful and intuitive tool for reconstructing particle size distributions from magnetization measurements. - Highlights: • A new method for the size distribution fit of magnetic nanoparticles is proposed. • Employed Kaczmarz algorithm does not need a priori input or eigenwert regularization. • The method is highly robust to noise contamination. • Size distributions are reconstructed from simulated and measured magnetization curves.

Signal processing of data from short sample tests for the projection of conductor performance in ITER magnets

International Nuclear Information System (INIS)

Martovetsky, Nicolai

2008-01-01

Qualification of the ITER conductor is absolutely necessary. Testing large scale conductors is expensive and time-consuming. To test 3-4 m long straight samples in a bore of a split solenoid is a relatively economical way in comparison with the fabrication of a coil to be tested in a bore of a background field solenoid. However, testing short samples may give ambiguous results due to different constraints in current redistribution in the cable or other end effects which are not present in the large magnet. This paper discusses the processes taking place in the ITER conductor, conditions when conductor performance could be distorted and possible signal processing to deduce the behaviour of ITER conductors in ITER magnets from the test data

Selection of a quench detection system for the ITER CS magnet

International Nuclear Information System (INIS)

Coatanea, Marc; Duchateau, Jean-Luc; Lacroix, Benoit; Nicollet, Sylvie; Rodriguez-Mateos, Felix; Topin, Frederic

2011-01-01

At variance with most of the existing superconducting systems operating in the world, the ITER central solenoid (CS) magnet is a fast pulsed system. This peculiarity creates a specific situation regarding the quench detection system, as a small resistive signal associated with a quench has to be discriminated from the high inductive signals imposed by the plasma scenario. The quench detection is based on an inductive compensation built from three adjacent double pancakes. The ITER protection rules for a superconducting magnet impose to respect the so-called maximum hot spot temperature criterion of 250 K in the quenched cable at the end of the fast discharge. A careful analysis of the residual inductive signals in the detection voltage shows that a blanking of the quench detection cannot be avoided during the early times of the plasma discharge (i.e. during 3.5 s). It is demonstrated that this blanking is, however, acceptable while fulfilling the hot spot criterion because the plasma initiation phase (PIP) is very similar to a fast safety discharge and corresponds to a fast decrease of the modules currents, which is favourable for the magnet protection.

Development of a magnetic diagnostic suitable for the ITER radiation environment

International Nuclear Information System (INIS)

Moreau, P.; Le-Luyer, A.; Malard, P.; Pastor, P.; Fournier, Y.; Lister, J. B.; Moret, J. M.; Testa, D.; Toussaint, M.; Chitarin, G.; Delogu, R.; Galo, A.; Peruzzo, S.; Romero, J.; Vila, R.; Brichard, B.; Bolshakova, I.; Duran, I.; Encheva, A.; Vayakis, G.

2009-01-01

Magnetic diagnostics of the ITER tokamak must fulfill demanding specifications, because their accuracy and reliability affects margins to the machine engineering limits and therefore operational flexibility. This paper describes the challenging issues related to the implementation of the magnetic diagnostics in a tokamak environment. We focus on nuclear radiations as they can significantly affect the measurement through Radiation Induced Electromotive Force (RIEMF) or Thermally Induced Electromotive Force (TIEMF). Thermal modeling of magnetic sensors and associated design studies are also reported as the thermal gradient in the sensors must be reduced to avoid TIEMF. Alternative magnetic sensors such as fiber optic current sensors (FOCS) or steady state magnetic field sensors are also discussed because they serve as a backup to the usual inductive magnetic measurements. We conclude by a brief review of the development needs for magnetic diagnostics. (authors)

The ITER magnets: Preparation for full size construction based on the results of the model coil programme

International Nuclear Information System (INIS)

Huguet, M.

2003-01-01

The ITER magnets are long-lead time items and the preparation of their construction is the subject of a major and coordinated effort of the ITER International Team and Participant Teams. The results of the ITER model coil programme constitute the basis and the main source of data for the preparation of the technical specifications for the procurement of the ITER magnets. A review of the salient results of the ITER model coil programme is given and the significance of these results for the preparation of full size industrial production is explained. The model coil programme has confirmed the validity of the design and the manufacturer's ability to produce the coils with the required quality level. The programme has also allowed the optimisation of the conductor design and the identification of further development which would lead to cost reductions of the toroidal field coil case. (author)

Nb3Sn conductor development for the ITER magnets

International Nuclear Information System (INIS)

Mitchell, N.

1997-01-01

The ITER magnet system consists of Toroidal Field (TF) coils, Poloidal Field (PF) coils, the Central Solenoid (CS) and error field correction coils (CC). The conductors for the coils are Nb 3 Sn or NbTi cable in conduit type, forced flow cooled with supercritical helium having a maximum operating current in the range 40-60 kA. To qualify the Nb 3 Sn conductor, two large model coils (energy up to 640 MJ) are being wound by the Home Teams of the Parties to the ITER EDA Agreement. A total of 24 t of strand has been completed for the CS model coil and 4 t for the TF model coil, and fabricated into 7 km of conductor in unit lengths up to 210 m, by an international collaboration involving 12 companies in Europe, Japan, Russia and the USA

Characterization of the radiation resistance of ITER-relevant and innovative fiber composites for the ITER magnet system

International Nuclear Information System (INIS)

Bittner-Rohrhofer, K.

2003-06-01

The application of glass-fiber reinforced composites for the insulation of the superconducting magnet coils of the ITER (International Thermonuclear Experimental Reactor ) fusion device requires high material performance. The mechanical integrity of the insulation is influenced by the neutron- and g-environment and by the high mechanical stresses of the magnet system over the entire plant lifetime of 20 years. Materials suggested as insulation have to be investigated in extensive test programs with respect to the present ITER design criteria. In particular, the ultimate tensile strength as well as the interlaminar shear behavior will change under static and dynamic load (tension-tension fatigue) at 77 K after irradiation to the ITER design fluence level of 1x1022 m-2 (E620.1 MeV). Therefore, a frequency of 10 Hz and a ratio of 0.1 were chosen, in order to simulate the pulsed TOKAMAK-operation as closely as possible. Furthermore, the fatigue behavior of the material is investigated over more than 3x104 cycles, which is the ITER- relevant design fatigue limit. Basically, these insulation systems are based on combined glass-fiber/Kapton tapes, which are impregnated with di-functional DGEBA epoxy resins. Several mechanical investigations showed that the radiation resistance of these organic resins is dramatically affected by radiation at a neutron fluence of 1x1022 m-2 (E620.1 MeV). Moreover, the material strength after irradiation is strongly influenced by these factors: the winding direction of the tapes, the quality of fabrication and the drastic delamination process of the whole compound. Furthermore, the radiation induced damage of adhesives applied for supporting the interfacial bonding between the glass-fiber tape and Kapton has an adverse effect on the material performance. In addition, the poor interlaminar shear behavior does not fulfil the requirements of ITER. These test-results motivated for the development of innovative resin systems with higher stability

Fatigue behavior of an insulation system for the ITER magnets

International Nuclear Information System (INIS)

Prokopec, R.; Humer, K.; Weber, H.W.

2006-01-01

The application of glass-fiber reinforced plastics as insulation materials for fusion magnet coils (e.g. the Toroidal Field Coils of ITER) requires the full characterization of their mechanical performance under ITER-relevant conditions. One of the methods of testing material's response under dynamic load is the tension-tension fatigue procedure. This test can be used to simulate the pulsed tokamak-operation of the ITER coils over a lifetime of more than 20 years. Furthermore, it provides information on the maximum tensile or shear stress in the ITER-relevant range of 10 4 -10 5 cycles. In order to simulate the operation conditions of ITER as closely as possible, several fatigue parameters can be set in the test programme, e.g., the minimum-to-peak stress ratio R and the frequency ν of the sinusoidal load function. Further, the fatigue process can be run under load or strain control. All of these parameters may influence the mechanical response of the insulation system under cyclic load. Therefore, it is highly desirable to investigate the influence of test parameter variations on the measured stress-lifetime diagrams. The investigations were performed at 77 K using an industrial glass-fiber reinforced composite impregnated with epoxy resin. For both the load and the strain controlled mode, R-values of 0.3 and 0.5 and a frequency of 10 Hz were chosen. The results showed almost no deviations in the lifetime behavior between the load and the strain controlled mode, up to the ITER specified number of pulses, i.e. 3 x 10 4 cycles. Beyond this point, the residual strength levels were lower by 5-30 % under strain control than under load control. This effect is more pronounced at higher cycle numbers and for lower R-ratios. (author)

Final design of the ITER outer vessel steady-state magnetic sensors.

Czech Academy of Sciences Publication Activity Database

Kocan, M.; Ďuran, Ivan; Entler, Slavomír; Vayakis, G.; Carmona, M.J.; Gitton, P.; Guirao, J.; Gonzalez, M.; Iglesias, S.; Pascual, Q.; Sandford, G.; Vacas, C.; Walsh, M.; Walton, R.

2017-01-01

Roč. 123, November (2017), s. 936-939 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : ITER * Magnetic sensor * Hall sensor Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S092037961730265X

Magnetic field compatibility of components for ITER electrical distribution systems

International Nuclear Information System (INIS)

Hourtoule, J.; Van Houtte, D.; Hertout, P.; Fejoz, P.

2004-01-01

The electrical distribution boards and control cubicles, installed inside the ITER Tokamak building, can be subject to a constant or slowly variable magnetic field up to 70 mT, 10 mT/s induced by ITER coils. This is a very unusual environmental condition and there are almost no data available on static magnetic field compatibility for the standard industrial electrical and electronic components that might be used inside standard low voltage distribution boards and standard control boards. CEA has especially taken in charge the test of electronic, control and signal conditioning units. For this purpose a test bed composed of a solenoid and a 30 V, 800 A power supply, has been developed at CEA Cadarache. Its characteristics are the following: - Magnetic field capability: 40 mT; - Variation of field: up to 10 mT/s. Useful dimensions for equipment under test: 500 x 500 x 500 mm. The list of the components to be tested has been identified trying to find common and recent components of different manufacturers. Test procedures have been written following the most relevant IEC standards and manufacturing recommendations and procedures. Magnetic compatibility tests on the basic components of standard low voltage control boards has been performed. All components tested are more or less sensitive to magnetic fields and the effect varies from the simple perturbation of the output signals to the partial destruction of some electronic card. The most sensitive are the ferromagnetic components like small power supplies transformers, relays, on inductors. For power supplies the most important consequence is the increase of the consumption (primary current). For most of the conditioning units the operational limits were found in the range of 30 mT (destruction of materials and overload of the supply) For PLC (Programme Logic Control) central units tested, the limit is in the order of 40 mT (components permanently out of service) Relays are limited to 15 mT (outputs oscillations, no

Test facility for joints of subsize cable-in-conduit conductors for NET/ITER winding studies

International Nuclear Information System (INIS)

Jager, B.; Chaussonnet, P.; Ciazynski, D.; Serries, J.P.; Simon, R.

1996-01-01

The joints between superconducting cables of the ITER coils present a major problem to be solved. A low resistance is required for the cooling, whereas small losses in variable magnetic field are needed for the stability. Different types of joints have already been suggested. Some preliminary measurements have already been made on prototypes of joints cooled in a helium bath at 4.2 K and for low currents. The test facility presented herein allows testing joint prototypes at the 1/8th scale (current up to 10 kA). These joints are cooled under the same conditions as those of ITER, by circulation of supercritical helium. A dipole allows submitting the joint to magnetic fields up to 4 teslas and to field variation rates up to dB/dt = 42 T/s

Design of the ITER magnets to provide plasma operational flexibility

International Nuclear Information System (INIS)

Mitchell, N.; Bessette, D.; Ferrari, M.; Huguet, M.; Jong, C.; Takahashi, Y.; Yoshida, K.; Maix, R.; Krivchenkov, Y.; Zapretilina, E.

2005-01-01

The ITER magnets have been optimised and refined since the ITER Final Design Report (FDR) in 2001. Multiple design options have been eliminated and there is improved ability to drive a wide range of plasma configurations. Design iterations on the TF out of plane supports have eliminated stress concentrations in the inner keyways and have led to the choice of a so called friction-joint on the outside. The closure procedure for the TF case has been changed, with a new case segmentation, less risk of winding pack damage from shrinkage and better filling of the case-winding gaps. Selection of compact joints for the CS has enabled the peak field and cyclic stress levels in the conductor to be reduced while maintaining the flux capability. The uncertainty in the nuclear heat levels in the inner legs of the TF coils, and the need to operate with plasma nuclear powers from 360 to 700MW, lead to a thermal screen on the inside of the case with variable cooling capability. The electrical insulation specification has been refined after irradiation test results to give a better margin on the onset of degradation after operation to 3MWa/m 2 . The RWM stabilisation provided by the side CC has been extended by accepting higher voltages and heating from AC losses. R and D results from the model coil tests have shown lower than expected design margins for the Nb3Sn conductors. This has been offset by adopting the latest advances in strand performance, and the margins of the new conductor will be confirmed by testing in 2005. Preparation for procurement is underway with considerations on technically acceptable ways of splitting the magnet supply. (author)

Preliminary optical design of polarization splitter box for ITER ECE diagnostic system

International Nuclear Information System (INIS)

Kumar, Ravinder; Danani, Suman; Pandya, Hitesh Kumar; Kumar, Vinay

2015-01-01

In tokamak, electron cyclotron emission (ECE) leaves the magnetically confined plasma with two polarizing modes, one with electric field parallel to magnetic field known as ordinary mode or O-Mode polarization, and other with the electric field perpendicular to magnetic field, extraordinary Mode or X-Mode. These radiation modes will be collected simultaneously in the ITER ECE measurement line. Therefore, it is necessary to split the radiation into O and X-mode polarizations before transmission otherwise there might be polarization mixing during transmission of the ECE radiation from tokamak to the measurement instruments. Proposed design of the polarization splitter box consists of two Gaussian beam telescopes built from three ellipsoidal mirrors and one flat mirror. A wire grid beam splitter separates the O and X-Mode polarization emission. The box is covered with microwave absorber to minimize scattering of the radiation. The design is being optimized by simulation using the Gaussian beam Mode software to achieve the desired performance, details will be discussed

Iterative reconstruction of magnetic induction using Lorentz transmission electron tomography

International Nuclear Information System (INIS)

Phatak, C.; Gürsoy, D.

2015-01-01

Intense ongoing research on complex nanomagnetic structures requires a fundamental understanding of the 3D magnetization and the stray fields around the nano-objects. 3D visualization of such fields offers the best way to achieve this. Lorentz transmission electron microscopy provides a suitable combination of high resolution and ability to quantitatively visualize the magnetization vectors using phase retrieval methods. In this paper, we present a formalism to represent the magnetic phase shift of electrons as a Radon transform of the magnetic induction of the sample. Using this formalism, we then present the application of common tomographic methods particularly the iterative methods, to reconstruct the 3D components of the vector field. We present an analysis of the effect of missing wedge and the limited angular sampling as well as reconstruction of complex 3D magnetization in a nanowire using simulations. - Highlights: • We present a formalism to represent electron-optical magnetic phase shift as a Radon transform of the 3D magnetic induction of the nano-object. • We have analyzed four different tomographic reconstruction methods for vectorial data reconstruction. • Reconstruction methods were tested for varying experimental limitations such as limited tilt range and limited angular sampling. • The analysis showed that Gridrec and SIRT methods performed better with lower errors than other reconstruction methods

Preliminary assessment of the tritium inventory and permeation in the plasma facing components of ITER

International Nuclear Information System (INIS)

Federici, G.; Holland, D.; Brooks, J.; Causey, R.; Dolan, T.J.; Longhurst, G.

1995-01-01

This paper discusses preliminary quantitative predictions for the tritium inventory in- and permeation through the first-wall and divertor PFC's of ITER. The primary plasma facing material under consideration is beryllium, with possible use of tungsten or carbon fiber composites (CFC's) on high-heat-flux surfaces. They use state-of-the-art tritium transport models, in conjunction with design parameters, and loading conditions anticipated for the first-wall, baffle, limiter and divertor. The analysis includes the synergistic effects of erosion on tritium implantation and trapping, which are expected to play a key role, particularly in the divertor regions where the interaction of the plasma with the surfaces will be most severe. The influence of several key parameters that strongly affect tritium build-up and release is assessed. Finally, they discuss the uncertainties in materials properties under ITER operating conditions and the R and D needed to resolve these uncertainties

ITER safety

International Nuclear Information System (INIS)

Raeder, J.; Piet, S.; Buende, R.

1991-01-01

As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

Magnetic Configuration Control of ITER Plasmas

International Nuclear Information System (INIS)

Albanese, R.; Artaserse, G.; Mattei, M.; Ambrosino, G.; Crisanti, F.; Tommasi, G. de; Fresa, R.; Portone, A.; Sartori, F.; Villone, F.

2006-01-01

The aim of this paper is to review the capability of the ITER Poloidal Field (PF) system of controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. The paper is broadly divided in two main sections devoted, respectively, to open loop (feed-forward) and closed loop (feedback) control. In the first part of the study the PF system is assessed with respect to the initiation, ramp-up, sustained burn, ramp-down phases of the main plasma inductive scenario. The limiter-to-divertor configuration transition phase is considered in detail with the aim of assessing the PF capability to form an X-point at the lowest possible current and, therefore, to relax the thermal load on the limiter surfaces. Moreover, during the sustained burn it is important to control plasmas with a broad range of current density profiles. In the second part of the study the plasma vertical feedback control requirements are assessed in details, in particular for the high elongation configurations achievable during the early limiter-to-X point transition phase. Non-rigid plasma displacement models are used to assess the control system voltage and current requirements of different radial field control circuits obtained, for example, by connecting the outermost PF coils, some CS coils, coils sub-sections etc. At last, the main 3D effects of the vessel ports are modeled and their impact of vertical stabilization evaluated. (author)

Preliminary ITER cost and schedule estimates

International Nuclear Information System (INIS)

1990-01-01

The cost, manpower requirements, and schedule estimates for the realization of the ITER tokamak have been studied during the Conceptual Design Activities, as a result of work by the ITER Management Committee. This work was completed during the January-March, 1990 joint work session, and is presented in this report. A possible schedule shows completion of the engineering design phase in 1995, with 180 professionals, at a cost of about $250M. The construction would be completed in 2004 with a rise in professional staff to 300, and a total cost of $4900M. The machine would be operable over an 18-year period, at an annual operating cost averaging $290M. 2 figs

Procurement of the ITER magnet components supplied by F4E

International Nuclear Information System (INIS)

Sborchia, Carlo; Oliva, Alessandro Bonito; Rajainmaeki, Hannu; Vostner, Alexander

2009-01-01

Fusion for Energy (F4E) will manage the in-kind contribution from Europe (EU) to ITER of about one quarter of the machine magnet components. This contribution will consist in 10 Toroidal Field (TF) coils, 5 (out of 6) Poloidal Field (PF) coils, 20% of the Nb3Sn conductor used in the TF coils and 15% of the PF coil NbTi conductor. In addition, 12 pre-compression fibreglass rings for the TF system will be procured. This paper presents the complex organization and status of the procurement activities of the F4E magnet components. The technical and managerial challenges, in terms of fabrication, quality assurance and schedule, are also highlighted.

Cryogenic Properties of Inorganic Insulation Materials for ITER Magnets: A Review

International Nuclear Information System (INIS)

Simon, N.J.

1994-01-01

Results of a literature search on the cryogenic properties of candidate inorganic insulators for the ITER TF magnets are reported. The materials investigated include: Al 2 O 3 , AlN, MgO, porcelain, SiO 2 , MgAl 2 O 4 , ZrO 2 , and mica. A graphical presentation is given of mechanical, elastic, electrical, and thermal properties between 4 and 300 K. A companion report reviews the low temperature irradiation resistance of these materials

Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Preliminary results

International Nuclear Information System (INIS)

Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

1993-01-01

Candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at temperatures of either 60 or 250 degrees C. Preliminary results have been obtained for several of these materials irradiated at 60 degrees C. The results show that irradiation at this temperature reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The unloading compliance technique developed for the subsize disk compact specimens works quite well, particularly for materials with lower toughness. Specimens of materials with very high toughness deform excessively, and this results in experimental difficulties

Fast ion power loads on ITER first wall structures in the presence of NTMs and microturbulence

International Nuclear Information System (INIS)

Kurki-Suonio, T.; Asunta, O.; Hirvijoki, E.; Koskela, T.; Snicker, A.; Sipilae, S.; Hauff, T.; Jenko, F.; Poli, E.

2011-01-01

The level and distribution of the wall power flux of energetic ions in ITER have to be known accurately in order to ensure the integrity of the first wall. Until now, most quantitative estimates have been based on the assumption that fast ion transport is dictated by neoclassical effects only. However, in ITER, the fast ion distribution is likely to be affected by various MHD effects and probably also by microturbulence. We have now upgraded our orbit-following Monte Carlo code ASCOT so that it has simple, theory-based models for neoclassical tearing mode (NTM)-type islands as well as for turbulent diffusion. ASCOT also allows for full-orbit following, which is important close to the material surfaces and, possibly, also when strong toroidal inhomogeneities are present in the magnetic field. Here we introduce the new models, preliminary results obtained with them, and how these models could be made more realistic in the future. The simulations are carried out for thermonuclear alpha particles in ITER scenario 2 plasma, because we consider this combination to be most critical for the successful operation of ITER. Neither the turbulent transport nor NTM-type islands are found to introduce alarming changes in the wall loads. However, at this stage it was not possible to combine the island structures with the non-axisymmetric magnetic field of ITER, and it remains to be seen what the combined effect of drift islands together with the toroidal ripple and local field aberrations, such as those due to test blanket modules and resonant magnetic perturbations will be.

Air-cooled fast discharge resistors for ITER magnets

International Nuclear Information System (INIS)

Tanchuk, Victor; Grigoriev, Sergey; Lokiev, Vladimir; Roshal, Alexander; Song, Inho; Buzykin, Oleg

2011-01-01

The ITER superconducting magnets will store up to 50 GJ of magnetic energy per operation cycle. In case of coil quench the energy stored in the coils must be extracted rapidly with a time constant from 7.5 to 14 s. It will be achieved by fast discharge resistors (FDR) normally bridged by circuit breakers and inserted in series with the superconducting coils. The fast discharge of the coils results practically in adiabatic heating of the resistive elements up to 200-300 deg. C. The resistors need to be cooled to the initial temperature over 6-8 h. Natural air circulation is proposed as a cooling method. In order to simulate the temperature response of the resistors to energy released in the resistive plates and to demonstrate their cooling capability within the required time by natural air circulation the numerical model of the resistor cooling circuit has been developed. As the calculations have shown, the developed FDR cooling system based on cooling by natural air circulation is capable of providing the required temperature operation regime of FDRs, but the supply channels are to be optimized so that the cooling time does not exceed the permissible one.

Cryogenic Properties of Inorganic Insulation Materials for ITER Magnets: A Review

Energy Technology Data Exchange (ETDEWEB)

Simon, N.J.

1994-12-01

Results of a literature search on the cryogenic properties of candidate inorganic insulators for the ITER TF magnets are reported. The materials investigated include: Al{sub 2}O{sub 3}, AlN, MgO, porcelain, SiO{sub 2}, MgAl{sub 2}O{sub 4}, ZrO{sub 2}, and mica. A graphical presentation is given of mechanical, elastic, electrical, and thermal properties between 4 and 300 K. A companion report reviews the low temperature irradiation resistance of these materials.

Fatigue behavior of an insulation system for the ITER magnets in the load and strain controlled mode

International Nuclear Information System (INIS)

Prokopec, R.; Humer, K.; Weber, H.W.

2007-01-01

The application of glass-fiber reinforced plastics as insulation materials for fusion magnet coils (e.g. of ITER) requires a full mechanical material characterization under ITER relevant conditions. The tension-tension fatigue test is useful to simulate the pulsed tokamak operation of the ITER coils in the relevant range of 10 4 -10 5 cycles. The fatigue process can be run under load or strain control, which may influence the material behavior under cyclic load conditions. Therefore, investigations were performed at 77 K using an industrial glass-fiber reinforced composite impregnated with epoxy resin. For both the load and the strain controlled mode, R-values of 0.3 and 0.5 and a frequency of 10 Hz were chosen. The results are discussed with respect to the lifetime performance of ITER

Preliminary RAMI analysis of DFLL TBS for ITER

Energy Technology Data Exchange (ETDEWEB)

Wang, Dagui [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230031 (China); Yuan, Run [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Jiaqun, E-mail: jiaqun.wang@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Fang; Wang, Jin [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

2016-11-15

Highlights: • We performed the functional analysis of the DFLL TBS. • We performed a failure mode analysis of the DFLL TBS. • We estimated the reliability and availability of the DFLL TBS. • The ITER RAMI approach was applied to the DFLL TBS for technical risk control in the design phase. - Abstract: ITER is the first fusion machine fully designed to prove the physics and technological basis for next fusion power plants. Among the main technical objectives of ITER is to test and validate design concepts of tritium breeding blankets relevant to the fusion power plants. To achieve this goal, China has proposed the dual functional lithium-lead test blanket module (DFLL TBM) concept design. The DFLL TBM and its associated ancillary system were called DFLL TBS. The DFLL TBS play a key role in next fusion reactor. In order to ensure reliable and available of DFLL TBS, the risk control project of DFLL TBS has been put on the schedule. As the stage of the ITER technical risk control policy, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was used to control the technical risk of ITER. In this paper, the RAMI approach was performed on the conceptual design of DFLL TBS. A functional breakdown was prepared on DFLL TBS, and the system was divided into 3 main functions and 72 basic functions. Based on the result of functional breakdown of DFLL TBS, the reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the DFLL TBS expected after implementation of mitigation actions was calculated to be 98.57% over 2 years based on the ITER reliability database. A Failure Modes Effects and Criticality Analysis (FMECA) was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability.

Alara applied to iter design and operation

International Nuclear Information System (INIS)

Uzan-Elbez, Joelle; Rodriguez-Rodrigo, Lina; Porfiri, Maria Teresa; Taylor, Neil; Gordon, Charles; Garin, Pascal; Girard, Jean-Philippe

2005-01-01

Based on the existing data on ITER and the safety options for licensing ITER in Cadarache, the present work assesses the application of the as-low-as-reasonably-achievable (ALARA) principle, as it has been implemented in the design of ITER and will be applied during ITER operation, as well as the compliance of the design with EUR/96-29 directive and regulation applicable in France. The preliminary occupational radiation exposure estimate gives a value of about 250 man mSv/a, which is half the annual target for ITER and comes essentially from maintenance activities. Some examples of the approach are presented

Development of Bismuth Hall sensors for ITER steady state magnetic diagnostics.

Czech Academy of Sciences Publication Activity Database

Ďuran, Ivan; Entler, Slavomír; Kočan, M.; Kohout, Michal; Viererbl, L.; Mušálek, Radek; Chráska, Tomáš; Vayakis, G.

2017-01-01

Roč. 123, November (2017), s. 690-694 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA MŠk LG14002 Institutional support: RVO:61389021 ; RVO:68378271 Keywords : ITER * Magnetic diagnostic * Hall sensor * Bismuth * Neutron irradiation * Radiation hardness Subject RIV: JF - Nuclear Energetics; JF - Nuclear Energetics (FZU-D) OBOR OECD: Nuclear related engineering; Nuclear related engineering (FZU-D) Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617306956

ITER Conceptual design: Interim report

International Nuclear Information System (INIS)

1990-01-01

This interim report describes the results of the International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activities after the first year of design following the selection of the ITER concept in the autumn of 1988. Using the concept definition as the basis for conceptual design, the Design Phase has been underway since October 1988, and will be completed at the end of 1990, at which time a final report will be issued. This interim report includes an executive summary of ITER activities, a description of the ITER device and facility, an operation and research program summary, and a description of the physics and engineering design bases. Included are preliminary cost estimates and schedule for completion of the project

Investigation of linearity of the ITER outer vessel steady-state magnetic field sensors at high temperature

Science.gov (United States)

Entler, S.; Duran, I.; Kocan, M.; Vayakis, G.

2017-07-01

Three vacuum vessel sectors in ITER will be instrumented by the outer vessel steady-state magnetic field sensors. Each sensor unit features a pair of metallic Hall sensors with a sensing layer made of bismuth to measure tangential and normal components of the local magnetic field. The influence of temperature and magnetic field on the Hall coefficient was tested for the temperature range from 25 to 250 oC and the magnetic field range from 0 to 0.5 T. A fit of the Hall coefficient normalized temperature function independent of magnetic field was found, and a model of the Hall coefficient functional dependence at a wide range of temperature and magnetic field was built with the purpose to simplify the calibration procedure.

Protection measures for selected ITER magnet system off-normal conditions

International Nuclear Information System (INIS)

Yoshida, K.; Iida, F.; Gallix, R.; Britousov, N.; Mitchell, N.; Thome, R.J.

1998-01-01

The International Thermonuclear Experimental Reactor (ITER) magnet systems provide the magnetic field intensity and field geometry to contain and control plasma during the various phases of pulsed operation. During these pulses, the toroidal field (TF) coils operate with a constant current. The central solenoid (CS) and poloidal field (PF) coils, on the other hand, are each independently powered. The maximum terminal voltages during plasma operation and protective discharges are 15 kV for CS and 10 kV for TF and PF. The energy stored in the 20 TF coil system is 103 GJ; in each of the other coils it is approximately 10 GJ or less. This paper describes the protection requirements and selected design concepts being considered for the large superconducting coils for the ITER. Ground faults, short circuits and helium leaks are the major serious accidents to be prevented in the coils. All coils use a solid insulation system to avoid ground faults. The electrical circuits including coil and power supply are grounded through resistors that limit current in the event of a ground fault. In the case of a short circuit within the coil winding, a large energy would be dissipated close to the small shorted volume. The impact of the short circuit can be reduced by using a potential screen. Inside the cryostat, helium leakage is most likely at the electrical insulating breaks in the cryogenic cooling lines between the coils and helium manifolds. A double containment (metallic shield and glass-epoxy) is therefore provided for the insulation breaks to allow for the detection of small leaks and to limit the spread of helium to other locations. (orig.)

Preliminary failure modes and effects analysis on Korean HCCR TBS to be tested in ITER

International Nuclear Information System (INIS)

Ahn, Mu-Young; Cho, Seungyon; Jin, Hyung Gon; Lee, Dong Won; Park, Yi-Hyun; Lee, Youngmin

2015-01-01

Highlights: • Postulated initiating events are identified through failure modes and effects analysis on the current HCCR TBS design. • A set of postulated initiating events are selected for consideration of deterministic analysis. • Accident evolutions on the selected postualted initiating events are qualitatively described for deterministic analysis. - Abstract: Korean Helium cooled ceramic reflector (HCCR) Test blanket system (TBS), which comprises Test blanket module (TBM) and ancillary systems in various locations of ITER building, is operated at high temperature and pressure with decay heat. Therefore, safety is utmost concern in design process and it is required to demonstrate that the HCCR TBS is designed to comply with the safety requirements and guidelines of ITER. Due to complexity of the system with many interfaces with ITER, a systematic approach is necessary for safety analysis. This paper presents preliminary failure modes and effects analysis (FMEA) study performed for the HCCR TBS. FMEA is a systematic methodology in which failure modes for components in the system and their consequences are studied from the bottom-up. Over eighty failure modes have been investigated on the HCCR TBS. The failure modes that have similar consequences are grouped as postulated initiating events (PIEs) and total seven reference accident scenarios are derived from FMEA study for deterministic accident analysis. Failure modes not covered here due to evolving design of the HCCR TBS and uncertainty in maintenance procedures will be studied further in near future.

Preliminary failure modes and effects analysis on Korean HCCR TBS to be tested in ITER

Energy Technology Data Exchange (ETDEWEB)

Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Jin, Hyung Gon; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Highlights: • Postulated initiating events are identified through failure modes and effects analysis on the current HCCR TBS design. • A set of postulated initiating events are selected for consideration of deterministic analysis. • Accident evolutions on the selected postualted initiating events are qualitatively described for deterministic analysis. - Abstract: Korean Helium cooled ceramic reflector (HCCR) Test blanket system (TBS), which comprises Test blanket module (TBM) and ancillary systems in various locations of ITER building, is operated at high temperature and pressure with decay heat. Therefore, safety is utmost concern in design process and it is required to demonstrate that the HCCR TBS is designed to comply with the safety requirements and guidelines of ITER. Due to complexity of the system with many interfaces with ITER, a systematic approach is necessary for safety analysis. This paper presents preliminary failure modes and effects analysis (FMEA) study performed for the HCCR TBS. FMEA is a systematic methodology in which failure modes for components in the system and their consequences are studied from the bottom-up. Over eighty failure modes have been investigated on the HCCR TBS. The failure modes that have similar consequences are grouped as postulated initiating events (PIEs) and total seven reference accident scenarios are derived from FMEA study for deterministic accident analysis. Failure modes not covered here due to evolving design of the HCCR TBS and uncertainty in maintenance procedures will be studied further in near future.

ITER safety and operational scenario

International Nuclear Information System (INIS)

Shimomura, Y.; Saji, G.

1998-01-01

The safety and environmental characteristics of ITER and its operational scenario are described. Fusion has built-in safety characteristics without depending on layers of safety protection systems. Safety considerations are integrated in the design by making use of the intrinsic safety characteristics of fusion adequate to the moderate hazard inventories. In addition to this, a systematic nuclear safety approach has been applied to the design of ITER. The safety assessment of the design shows how ITER will safely accommodate uncertainties, flexibility of plasma operations, and experimental components, which is fundamental in ITER, the first experimental fusion reactor. The operation of ITER will progress step by step from hydrogen plasma operation with low plasma current, low magnetic field, short pulse and low duty factor without fusion power to deuterium-tritium plasma operation with full plasma current, full magnetic field, long pulse and high duty factor with full fusion power. In each step, characteristics of plasma and optimization of plasma operation will be studied which will significantly reduce uncertainties and frequency/severity of plasma transient events in the next step. This approach enhances reliability of ITER operation. (orig.)

ITER ITA newsletter. No. 11, December 2003

International Nuclear Information System (INIS)

2003-12-01

This issue of the ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER including information from the editor about ITER update, about progress in ITER magnet design and preparation of procurement packages and about 25th anniversary of the First Steering Committee Meeting of the International Tokamak Reactor (INTOR) Workshop, organized under the auspices of the IAEA, took place at the IAEA Headquarters in Vienna

ITER ITA newsletter. No. 10, November 2003

International Nuclear Information System (INIS)

2003-12-01

This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about an ITER related meeting, namely, the Ninth ITER Negotiations Meeting (N-9), which was held on 9-10 November 2003 at the Fragrant Hill Golden Resources Commerce Hotel in Beijing and information about research on magnetic confinement fusion (MCF) in China

Concept of magnet systems for LHD-type reactor

International Nuclear Information System (INIS)

Imagawa, S.; Takahata, K.; Tamura, H.; Yanagi, N.; Mito, T.; Obana, T.; Sagara, A.

2008-10-01

Heliotron reactors have attractive features for fusion power plants, such as no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, the major radius of plasma of 14 to 17 m with a central toroidal field of 6 to 4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The stored magnetic energy is estimated at 120 to 140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress can be comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than 1,000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is longer than 150 m that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with small extension of the ITER technology. (author)

ITER EDA newsletter. V. 8, no. 9

International Nuclear Information System (INIS)

1999-09-01

This edition of the ITER EDA Newsletter contains a contribution by the ITER Director, R. Aymar, on the subject of developments in ITER Physics R and D report on the completion of the ITER central solenoid model coils installation by H. Tsuji, Head fo the Superconducting Magnet Laboratory at JAERI in Naka, Japan. Individual abstracts are prepared for each of the two articles

ITER EDA Newsletter. V. 4, no. 5

International Nuclear Information System (INIS)

1995-05-01

This issue of the ITER EDA (Engineering Design Activities) Newsletter contains comments on the ITER project by the Permanent Representative of the Russian Federation to the International Organizations in Vienna; a report on the ITER Magnet Technical Meeting held at the Joint Work Site at Naka, Japan, April 19-21, 1995; and a contribution entitled ''ITER spouses cross the cultures''

Progress in the Design and Testing of In-Vessel Magnetic Pickup Coils for ITER

Czech Academy of Sciences Publication Activity Database

Peruzzo, S.; Brombin, M.; Palumbo, M.F.; Gonzalez, W.; Marconato, N.; Rizzolo, A.; Arshad, S.; Ma, Y.; Vayakis, G.; Suarez, A.; Ďuran, Ivan; Viererbl, L.; Lahodová, Z.

2016-01-01

Roč. 44, č. 9 (2016), s. 1704-1710 ISSN 0093-3813. [Symposium on Fusion Engineering (SOFE) colocated with the 20th Pulsed Power Conference/26./. Austin, 31.05.2015-04.06.2015] Institutional support: RVO:61389021 Keywords : Low-temperature cofired ceramic (LTCC) * magnetic diagnostics * mineral insulated cable (MIC) * ITER Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.052, year: 2016

Preliminary study of magnet design for an SSC

International Nuclear Information System (INIS)

Taylor, C.E.; Meuser, R.B.

1983-08-01

The overriding design consideration for the SSC magnets is that cost of the facility be minimized; at 8 T, approximately 40 km of bending magnets is required for each ring of a 20 TeV collider. We present some results of a parametric study of two-in-one, iron-core magnets for an SSC. These results are necessarily preliminary in nature, and are intended only to show some of the trade-offs for a wide range of the variables. We show also some results for a reference design that produces 6.5 T in the aperture at 4.4 K for a coil inside diameter of 40 mm. It is not to be inferred that we have established this to be an optimum in any sense

ITER EDA newsletter. V. 2, no. 12

International Nuclear Information System (INIS)

1993-12-01

This issue of the ITER EDA (Engineering Design Activities) Newsletter contains a report of the Second ITER Technical Committee Meeting on Safety, Environment, and Regulatory Approval, San Diego, USA, November 3-12, 1993, and a summary report on an ITER Magnet Technical Meeting, Naka, Japan, October 5-8, 1993

RAMI analysis for ITER radial X-ray camera system

Energy Technology Data Exchange (ETDEWEB)

Qin, Shijun, E-mail: sjqin@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Liqun; Chen, Kaiyun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Barnsley, Robin; Sirinelli, Antoine [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Song, Yuntao; Lu, Kun; Yao, Damao; Chen, Yebin; Li, Shi; Cao, Hongrui; Yu, Hong; Sheng, Xiuli [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

2016-11-15

Highlights: • The functional analysis of the ITER RXC system was performed. • A failure modes, effects and criticality analysis of the ITER RXC system was performed. • The reliability and availability of the ITER RXC system and its main functions were calculated. • The ITER RAMI approach was applied to the ITER RXC system for technical risk control in the preliminary design phase. - Abstract: ITER is the first international experimental nuclear fusion device. In the project, the RAMI approach (reliability, availability, maintainability and inspectability) has been adopted for technical risk control to mitigate all the possible failure of components in preparation for operation and maintenance. RAMI analysis of the ITER Radial X-ray Camera diagnostic (RXC) system during preliminary design phase was required, which insures the system with a very high performance to measure the X-ray emission and research the MHD of plasma with high accuracy on the ITER machine. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 3 main functions, 6 intermediate functions and 28 basic functions which are described using the IDEFØ method. Reliability block diagrams (RBDs) were prepared to calculate the reliability and availability of each function under assumption of operating conditions and failure data. Initial and expected scenarios were analyzed to define risk-mitigation actions. The initial availability of RXC system was 92.93%, while after optimization the expected availability was 95.23% over 11,520 h (approx. 16 months) which corresponds to ITER typical operation cycle. A Failure Modes, Effects and Criticality Analysis (FMECA) was performed to the system initial risk. Criticality charts highlight the risks of the different failure modes with regard to the probability of their occurrence and impact on operations. There are 28 risks for the initial state, including 8 major risks. No major risk remains after taking into

Novel quench detection methods for the superconducting magnets in ITER and TPX

International Nuclear Information System (INIS)

Schultz, J.H.; Pourrahimi, S.; Diatchenko, N.; Guss, W.; Chaniotakis, E.; Pillsbury, R.D. Jr.; Smith, S.; Wang, P.W.; Citrolo, J.; Chaplin, M.; Zbasnik, J.

1995-01-01

The US is providing novel sensors to Japan to be used in the conductor for QUELL, the ITER Quench Experiment on Long-Lengths to be performed in the SULTAN magnet in 1995. These include cowound voltage sensors, fiber optic thermometers, cowound and conventional pressure sensors, and flow meters. TPX has a redundant quench detection system using cowound voltage sensors, fiber-optic temperaure sensors, conventional voltage taps, and flow meters. Sensors are extracted only at joint regions, but are terminated every two pancakes, providing high signal-noise ratios through differencing techniques. (orig.)

Confinement margins for ignition and driven operation in Iter Eda ID

International Nuclear Information System (INIS)

Johner, J.

1995-09-01

Preliminary calculations for ITER EDA ID have been performed using the 1/2D thermal equilibrium code HELIOS. It is found that: - The maximum ignition margin for ITER ID (29%) is 6% less than for ITER OD (35%) and 5% less than for ITER CDA (34%). - Decreasing the ration τ * He /τ E from the nominal value 10 to a value of 5 gives a 12% gain in the maximum ignition margin. Increasing the ration from 10 to 15 causes a 22% loss in the margin. Furthermore, ignited equilibria non longer exist for τ * He /τ E ≥ 17.6. - Operation in driven mode with 50 MW of external power increases the confinement capability by 13%. With 100 MW, the improvement is 24%. - Lowering the fusion power from 1500 to 1000 MW slightly improves the maximum ignition margin (+5%) and allows operation below the Greenwald density limit. - A 10% reduction of the toroidal magnetic field with a correlative diminution of the plasma current for constant safety factor operation, causes a dramatic reduction (-18%) of the maximum ignition margin. - A fraction of neon of 0.68% would completely suppress the ignition margin. Furthermore, ignited equilibria, with the nominal fusion power and τ * He /τ E , no longer exist when the neon fraction exceeds 0.75%. (Author). 2 refs., 10 figs

Another year of successful collaboration between ITER and CERN

CERN Multimedia

Lucio Rossi and Frederick Bordry

The 4th meeting of the Steering Committee of the CERN-ITER Collaboration Agreement was held on 19 November at CERN. It marked the end of a second year of successful collaboration between ITER and CERN on superconducting magnets and associated technologies and the establishment of CERN as the ITER “reference laboratory” for superconducting strand testing for the next five years.   From left to right: Luca Bottura, Head of the CERN Superconductor and Devices Section, Neil Mitchell, Head of the ITER Magnet Division, Frederick Bordry, Head of the CERN Technology Department, Arnaud Devred, Head of the ITER Superconductor Systems and Auxiliaries Section, and Lucio Rossi, Head of the CERN Magnet, Superconductors and Cryostats Group, standing in front of a sample holder used for critical current measurements of Nb3Sn strands in the Superconductor Laboratory (bldg 163) at CERN. The implementation agreement for 2009 encompassed a wide range of topics ranging from expertise in stainless ...

ITER EDA newsletter. V. 2, no. 9

International Nuclear Information System (INIS)

1993-09-01

This ITER EDA (Engineering Design Activities) Newsletter issue contains a report on the third meeting of the ITER Technical Advisory Committee, a summary report for the ITER Magnetic Technical Meeting, a brief account of the International Workshop on Nuclear Data for Fusion Reactor Technology, and a description of approved arrangements for visiting home team personnel

ITER jako živý

Czech Academy of Sciences Publication Activity Database

Řípa, Milan

2010-01-01

Roč. 10, č. 6 (2010), s. 18-19 Institutional research plan: CEZ:AV0Z20430508 Keywords : Fusion * ITER * magnetic field * ELMs * cryo pumps * central solenoid * correction coils * superconducting coils * toroidal field coils * poloidal field coils * divertor * Cadarache Subject RIV: BL - Plasma and Gas Discharge Physics http://www.tretipol.cz/900-iter-jako-zivy

ITER EDA Newsletter. V.3, no.4

International Nuclear Information System (INIS)

1994-04-01

This ITER EDA Newsletter issue contains a report on the fifth meeting of the ITER Management Advisory Committee and a summary of a magnet and safety technical meeting held at Naka, February 22-25, 1994

Magnetic compatibility of standard components for electrical installations: Computation of the background field and consequences on the design of the electrical distribution boards and control boards for the ITER Tokamak building

International Nuclear Information System (INIS)

Benfatto, I.; Bettini, P.; Cavinato, M.; Lorenzi, A. De; Hourtoule, J.; Serra, E.

2005-01-01

Inside the proposed Tokamak building, the ITER poloidal field magnet system would produce a stray magnetic field up to 70 mT. This is a very unusual environmental condition for electrical installation equipment and limited information is available on the magnetic compatibility of standard components for electrical distribution boards and control boards. Because this information is a necessary input for the design of the electrical installation inside the proposed ITER Tokamak building specific investigations have been carried out by the ITER European Participant Team. The paper reports on the computation of the background magnetic field map inside the ITER Tokamak building and the consequences on the design of the electrical installations of this building. The effects of the steel inside the building structure and the feasibility of magnetic shields for electrical distribution boards and control boards are also reported in the paper. The results of the test campaigns on the magnetic field compatibility of standard components for electrical distribution boards and control boards are reported in companion papers published in these proceedings

A Regularized Approach for Solving Magnetic Differential Equations and a Revised Iterative Equilibrium Algorithm

International Nuclear Information System (INIS)

Hudson, S.R.

2010-01-01

A method for approximately solving magnetic differential equations is described. The approach is to include a small diffusion term to the equation, which regularizes the linear operator to be inverted. The extra term allows a 'source-correction' term to be defined, which is generally required in order to satisfy the solvability conditions. The approach is described in the context of computing the pressure and parallel currents in the iterative approach for computing magnetohydrodynamic equilibria.

ITER EDA and technology

International Nuclear Information System (INIS)

Baker, C.C.

2001-01-01

The year 1998 was the culmination of the six-year Engineering Design Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project. The EDA results in design and validating technology R and D, plus the associated effort in voluntary physics research, is a significant achievement and major milestone in the history of magnetic fusion energy development. Consequently, the ITER EDA was a major theme at this Conference, contributing almost 40 papers

Updated safety analysis of ITER

Energy Technology Data Exchange (ETDEWEB)

Taylor, Neill, E-mail: neill.taylor@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

2011-10-15

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

Updated safety analysis of ITER

International Nuclear Information System (INIS)

Taylor, Neill; Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid

2011-01-01

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

Preliminary analysis of patent trends for magnetic fusion technology

International Nuclear Information System (INIS)

Levine, L.O.; Ashton, W.B.; Campbell, R.S.

1984-02-01

This study presents a preliminary analysis of development trends in magnetic fusion technology based on data from US patents. The research is limited to identification and description of general patent activity and ownership characteristics for 373 patents. The results suggest that more detailed studies of fusion patents could provide useful R and D planning information

Structural analysis of the ITER Vacuum Vessel regarding 2012 ITER Project-Level Loads

Energy Technology Data Exchange (ETDEWEB)

Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Jun, C.H.; Portafaix, C.; Choi, C.-H.; Ioki, K.; Sannazzaro, G.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Cambazar, M.; Corti, Ph.; Pinori, K.; Sfarni, S.; Tailhardat, O. [Assystem EOS, 117 rue Jacquard, L' Atrium, 84120 Pertuis (France); Borrelly, S. [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Albin, V.; Pelletier, N. [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France)

2014-10-15

Highlights: • ITER Vacuum Vessel is a part of the first barrier to confine the plasma. • ITER Vacuum Vessel as Nuclear Pressure Equipment (NPE) necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, conformance testing and quality assurance, i.e. Agreed Notified Body (ANB). • A revision of the ITER Project-Level Load Specification was implemented in April 2012. • ITER Vacuum Vessel Loads (seismic, pressure, thermal and electromagnetic loads) were summarized. • ITER Vacuum Vessel Structural Margins with regards to RCC-MR code were summarized. - Abstract: A revision of the ITER Project-Level Load Specification (to be used for all systems of the ITER machine) was implemented in April 2012. This revision supports ITER's licensing by accommodating requests from the French regulator to maintain consistency with the plasma physics database and our present understanding of plasma transients and electro-magnetic (EM) loads, to investigate the possibility of removing unnecessary conservatism in the load requirements and to review the list and definition of incidental cases. The purpose of this paper is to present the impact of this 2012 revision of the ITER Project-Level Load Specification (LS) on the ITER Vacuum Vessel (VV) loads and the main structural margins required by the applicable French code, RCC-MR.

ITER EDA Newsletter. V. 3, no. 7

International Nuclear Information System (INIS)

1994-07-01

This ITER EDA (Engineering Design Activities) Newsletter issue reports on the sixth meeting on the ITER management advisory committee (MAC-6); on the sixth meeting of the ITER technical advisory committee (TAC-6); a summary of a magnet technical meeting, held at Naka, Japan, June 27-30 1994 is also included. It finally contains an in memoriam on the passing away of Dr. A.I. Kostenko

Key engineering features of the ITER-FEAT magnet system and implications for the R and D programme

International Nuclear Information System (INIS)

Huguet, M.

2001-01-01

The magnet design of the new ITER-FEAT machine comprises 18 Toroidal Field (TF) coils, a Central Solenoid (CS), 6 Poloidal Field (PF) coils and Correction Coils (CCs). A key driver of this new design is the requirement to generate and control plasmas with a relatively high elongation (k 95 =1.7) and a relatively high triangularity (δ 95 =0.35). This has lead to a design where the CS is vertically segmented and self-standing and the TF coils are wedged along their inboard legs. Another important design driver is to achieve a high operational reliability of the magnets, and this has resulted in several unconventional designs, and in particular, the use of conductors supported in radial plates for the winding pack of the TF coils. A key mechanical issue is the cyclic loading of the TF coil cases due to the out-of-plane loads which result from the interaction of the TF coil current and the poloidal field. These loads are resisted by a combination of shear keys and 'pre-compression' rings able to provide a centripetal preload at assembly. The fatigue life of the CS conductor jacket is another issue as it determines the CS performance in terms of the flux generation. Two jacket materials and designs are under study. Since 1993, the ITER magnet R and D programme has been focussed on the manufacture and testing of a CS and a TF model coil. During its testing, the CS model coil has successfully achieved all its performance targets in DC and AC operations. The manufacture of the TF model coil is complete. The manufacture of segments of the full scale TF coil case is another important and successful part of this programme and is near completion. New R and D effort is now being initiated to cover specific aspects of the ITER-FEAT design. (author)

Structural analysis of the Passive Magnetic Shield for the ITER Heating Neutral Beam Injector system

Energy Technology Data Exchange (ETDEWEB)

Cabrera, Santiago, E-mail: santiago.cabrera@ciemat.es [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); Rincón, Esther; Ahedo, Begoña; Alonso, Javier; Barrera, Germán; Ramos, Francisco; Ríos, Luis [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); El-Ouazzani, Anass; García, Pablo [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3 – 07/08, 08019 Barcelona (Spain)

2015-10-15

The ITER Passive Magnetic Shield (PMS) main function is to protect the Neutral Beam Injector (NBI) from the external magnetic field coming from the tokamak, and to shield the NB cell from the radiation coming from all activated components. The shielding from the external magnetic field is performed in association with the Active Compensation Cooled Correction Coils (ACCC). The Bushing and Transmission Line (TL) PMS also provides structural support for HV bushing, allowing its maintenance and providing air sealing function between NBI cell and High Voltage deck room. The paper summarizes the structural analyses performed in order to evaluate the mechanical behaviour of the HNB PMS under operation combined with seismic event. The RCC-MR Code is used to validate the design, assuming creep is negligible, since the structure is expected to be at room temperature. P-type damage is assessed.

Overview of physics basis for ITER

International Nuclear Information System (INIS)

Mukhovatov, V; Shimada, M; Chudnovskiy, A N; Costley, A E; Gribov, Y; Federici, G; Kardaun, O; Kukushkin, A S; Polevoi, A; Pustovitov, V D; Shimomura, Y; Sugie, T; Sugihara, M; Vayakis, G

2003-01-01

ITER will be the first magnetic confinement device with burning DT plasma and fusion power of about 0.5 GW. Parameters of ITER plasma have been predicted using methodologies summarized in the ITER Physics Basis (1999 Nucl. Fusion 39 2175). During the past few years, new results have been obtained that substantiate confidence in achieving Q>=10 in ITER with inductive H-mode operation. These include achievement of a good H-mode confinement near the Greenwald density at high triangularity of the plasma cross section; improvements in theory-based confinement projections for the core plasma, even though further studies are needed for understanding the transport near the plasma edge; improvement in helium ash removal due to the elastic collisions of He atoms with D/T ions in the divertor predicted by modelling; demonstration of feedback control of neoclassical tearing modes and resultant improvement in the achievable beta values; better understanding of edge localized mode (ELM) physics and development of ELM mitigation techniques; and demonstration of mitigation of plasma disruptions. ITER will have a flexibility to operate also in steady-state and intermediate (hybrid) regimes. The 'advanced tokamak' regimes with weak or negative central magnetic shear and internal transport barriers are considered as potential scenarios for steady-state operation. The paper concentrates on inductively driven plasma performance and discusses requirements for steady-state operation in ITER

Preliminary experimental results from a linear reciprocating magnetic refrigerator prototype

International Nuclear Information System (INIS)

Tagliafico, Luca Antonio; Scarpa, Federico; Valsuani, Federico; Tagliafico, Giulio

2013-01-01

A linear reciprocating magnetic refrigerator prototype was designed and built with the aid of an industrial partner. The refrigerator is based on the Active Magnetic Regenerative cycle, and exploits two regenerators working in parallel. The active material is Gadolinium in plates, 0.8 mm thick, for a total mass of 0.36 kg. The device is described and results about magnetic field and temperature span measurements are presented. The designed permanent magnet structure, based on an improved cross-type arrangement, generates a maximum magnetic field intensity of 1.55 T in air, over a gap of (13 × 50 × 100) mm 3 . The maximum temperature span achieved is 5.0 K, in a free run condition. -- Highlights: ► We give preliminary results from a linear reciprocating magnetic refrigerator prototype. ► The design is intended to process visualization and investigation. ► The prototype behavior gives us various suggestions to improve its general performance

Room Temperature Magnetic Determination of the Current Center Line for the ITER TF Coils

CERN Document Server

Lerch, Philippe; Buzio, Marco; Negrazus, Marco; Baynham, Elwyn; Sanfilippo, Stephane; Foussat, Arnaud

2014-01-01

The ITER tokamak includes 18 superconducting D-shaped toroidal field (IT) coils. Unavoidable shape deformations as well as assembly errors will lead to field errors, which can be modeled with the knowledge of the current center line (CCL). Accurate survey during the entire manufacturing and assembly process, including transfer of survey points, is complex. In order to increase the level of confidence, a room temperature magnetic measurement of the CCL on assembled and closed winding packs is foreseen, prior to insertion into their cold case. In this contribution, we discuss the principle of the CCL determination and present a low frequency ac measurement system under development at PSI, within an ITER framework contract. The largest current allowed to flow in the TF coil at room temperature and the precision requirements for the determination of the CCL loci of the coil are hard boundaries. Eddy currents in the radial plates, the winding pack enclosures, and possibly from iron in the reinforced concrete floor...

Fire risk analysis in ITER tritium building

International Nuclear Information System (INIS)

Lignini, Franck; Uzan-Elbez, Joelle; Girard, Jean-Philippe; Porfiri, Maria Teresa; Rodriguez-Rodrigo, Lina

2005-01-01

Events, such as fire, have been considered in ITER documentation of low probability and a general approach has been defined in [Technical basis for the ITER final design, EDA Documentation Series I, No. 22, IAEA, Vienna, 2001] to be developed later for the ITER specific site. It was said that 'these hazards will be treated according to the industrial safety regulations and practices of the host country'. In the framework of studies for the European ITER site in Cadarache, an assessment of fire hazard has been done in order to ensure compliance with French safety requirements. In this report, a summary of existing laws is presented and an example of the deterministic approach to be followed for the preliminary safety report (PSR) is given on the analysis of tritium building design

Radiation Limits for Nb3Sn Superconductors for ITER Magnets: A literature review

International Nuclear Information System (INIS)

Simon, N.J.

1995-01-01

The data base on radiation damage to Nb 3 Sn superconductors is compiled from the literature and assessed in this report. Nb 3 Sn superconductors are currently under procurement for use in ITER magnet prototypes. In contrast to the data base on insulation materials proposed for use in ITER magnets, the data base on the radiation damage of Nb 3 Sn is much more complete. Key results have often been confirmed by several groups at different institutions. The investigation of variables that influence radiation damage has also been much more complete for Nb 3 Sn than for insulators. Furthermore, in situ testing of superconducting parameters is much easier than in situ mechanical testing of insulators, and in situ testing has invariably been performed after cryogenic irradiation of Nb 3 Sn. However, in recent years, Nb 3 Sn testing has also suffered from the lack of 4-K irradiation facilities. Just as new processing methods to obtain more economical Nb 3 Sn conductor products in large quantity were being developed, cryogenic irradiation sources were being phased out. Therefore, this brief introductory section presents some basic information on the properties and structure of Nb 3 Sn superconducting composites and the distinctions between different fabrication processes. This provides a background to assess the adequacy of the current cryogenic data base on radiation damage, Also, since synergistic effects of strain and irradiation have recently been investigated, a brief discussion of the effects of strain on Nb 3 Sn properties is included in this introduction

A preliminary conceptual design study for Korean fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Kim, Hyoung Chan; Oh, Sangjun; Lee, Young Seok; Yeom, Jun Ho; Im, Kihak; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Kessel, Charles; Brown, Thomas; Titus, Peter [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States)

2013-10-15

Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb{sub 3}Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

Challenges and status of ITER conductor production

International Nuclear Information System (INIS)

Devred, A; Backbier, I; Bessette, D; Bevillard, G; Gardner, M; Jong, C; Lillaz, F; Mitchell, N; Romano, G; Vostner, A

2014-01-01

Taking the relay of the large Hadron collider (LHC) at CERN, ITER has become the largest project in applied superconductivity. In addition to its technical complexity, ITER is also a management challenge as it relies on an unprecedented collaboration of seven partners, representing more than half of the world population, who provide 90% of the components as in-kind contributions. The ITER magnet system is one of the most sophisticated superconducting magnet systems ever designed, with an enormous stored energy of 51 GJ. It involves six of the ITER partners. The coils are wound from cable-in-conduit conductors (CICCs) made up of superconducting and copper strands assembled into a multistage cable, inserted into a conduit of butt-welded austenitic steel tubes. The conductors for the toroidal field (TF) and central solenoid (CS) coils require about 600 t of Nb 3 Sn strands while the poloidal field (PF) and correction coil (CC) and busbar conductors need around 275 t of Nb–Ti strands. The required amount of Nb 3 Sn strands far exceeds pre-existing industrial capacity and has called for a significant worldwide production scale up. The TF conductors are the first ITER components to be mass produced and are more than 50% complete. During its life time, the CS coil will have to sustain several tens of thousands of electromagnetic (EM) cycles to high current and field conditions, way beyond anything a large Nb 3 Sn coil has ever experienced. Following a comprehensive R and D program, a technical solution has been found for the CS conductor, which ensures stable performance versus EM and thermal cycling. Productions of PF, CC and busbar conductors are also underway. After an introduction to the ITER project and magnet system, we describe the ITER conductor procurements and the quality assurance/quality control programs that have been implemented to ensure production uniformity across numerous suppliers. Then, we provide examples of technical challenges that have been

Challenges and status of ITER conductor production

Science.gov (United States)

Devred, A.; Backbier, I.; Bessette, D.; Bevillard, G.; Gardner, M.; Jong, C.; Lillaz, F.; Mitchell, N.; Romano, G.; Vostner, A.

2014-04-01

Taking the relay of the large Hadron collider (LHC) at CERN, ITER has become the largest project in applied superconductivity. In addition to its technical complexity, ITER is also a management challenge as it relies on an unprecedented collaboration of seven partners, representing more than half of the world population, who provide 90% of the components as in-kind contributions. The ITER magnet system is one of the most sophisticated superconducting magnet systems ever designed, with an enormous stored energy of 51 GJ. It involves six of the ITER partners. The coils are wound from cable-in-conduit conductors (CICCs) made up of superconducting and copper strands assembled into a multistage cable, inserted into a conduit of butt-welded austenitic steel tubes. The conductors for the toroidal field (TF) and central solenoid (CS) coils require about 600 t of Nb3Sn strands while the poloidal field (PF) and correction coil (CC) and busbar conductors need around 275 t of Nb-Ti strands. The required amount of Nb3Sn strands far exceeds pre-existing industrial capacity and has called for a significant worldwide production scale up. The TF conductors are the first ITER components to be mass produced and are more than 50% complete. During its life time, the CS coil will have to sustain several tens of thousands of electromagnetic (EM) cycles to high current and field conditions, way beyond anything a large Nb3Sn coil has ever experienced. Following a comprehensive R&D program, a technical solution has been found for the CS conductor, which ensures stable performance versus EM and thermal cycling. Productions of PF, CC and busbar conductors are also underway. After an introduction to the ITER project and magnet system, we describe the ITER conductor procurements and the quality assurance/quality control programs that have been implemented to ensure production uniformity across numerous suppliers. Then, we provide examples of technical challenges that have been encountered and

JSME construction standard for superconducting magnets of fusion facilities. Toward the construction of ITER

International Nuclear Information System (INIS)

Nakasone, Yuji; Takahashi, Yukio; Sato, Kazuyoshi; Nishimura, Arata; Suzuki, Tetsuya; Irie, Hirosada; Nakahira, Masataka

2009-01-01

The present paper describes the general view of the construction standard, which the Japan Society of Mechanical Engineers (JSME) has recently set up and published, for superconducting magnet structures to be used in nuclear fusion facilities. The present target of the standard is tokamak-type fusion energy facilities, especially the International Thermonuclear Experimental Reactor called ITER for short. The standard contains rules for structural materials including cryogenic materials, structural design considering magnetic forces, manufacture including welding and installation, nondestructive testing, pressure proof tests and leak tests of toroidal field magnet structures. The standard covers requirements for structural integrity, deformation control, and leak tightness of all the components of the superconducting magnets and their supports except for superconducting strands and electrical insulators. The standard does not cover deterioration, which may occur in service as a result of corrosion, radiation effects, or instability of material. The standard consists of seven articles and twelve mandatory and non-mandatory appendices to the articles; i.e., (1) Scope, roles and responsibilities, (2) Materials, (3) Structural design, (4) Fabrication and installation, (5) Non-destructive examination, (6) Pressure and leak testing, and (7) Terms used in general requirements. (author)

Pivotal issues on relativistic electrons in ITER

Science.gov (United States)

Boozer, Allen H.

2018-03-01

The transfer of the plasma current from thermal to relativistic electrons is a threat to ITER achieving its mission. This danger is significantly greater in the nuclear than in the non-nuclear phase of ITER operations. Two issues are pivotal. The first is the extent and duration of magnetic surface breaking in conjunction with the thermal quenches. The second is the exponential sensitivity of the current transfer to three quantities: (1) the poloidal flux change required to e-fold the number of relativistic electrons, (2) the time τa after the beginning of the thermal quench before the accelerating electric field exceeds the Connor-Hastie field for runaway, and (3) the duration of the period τ_op in which magnetic surfaces remain open. Adequate knowledge does not exist to devise a reliable strategy for the protection of ITER. Uncertainties are sufficiently large that a transfer of neither a negligible nor the full plasma current to relativistic electrons can be ruled out during the non-nuclear phase of ITER. Tritium decay can provide a sufficiently strong seed for a dangerous relativistic-electron current even if τa and τ_op are sufficiently long to avoid relativistic electrons during non-nuclear operations. The breakup of magnetic surfaces that is associated with thermal quenches occurs on a time scale associated with fast magnetic reconnection, which means reconnection at an Alfvénic rather than a resistive rate. Alfvénic reconnection is well beyond the capabilities of existing computational tools for tokamaks, but its effects can be studied using its property of conserving magnetic helicity. Although the dangers to ITER from relativistic electrons have been known for twenty years, the critical issues have not been defined with sufficient precision to formulate an effective research program. Studies are particularly needed on plasma behavior in existing tokamaks during thermal quenches, behavior which could be clarified using methods developed here.

Physics of thermo-nuclear fusion and the ITER project; La physique de la fusion thermonucleaire et le projet ITER

Energy Technology Data Exchange (ETDEWEB)

Garin, P [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee - DRFC, 13 - Saint-Paul-lez-Durance (France)

2003-01-01

This document gathers the slides of the 6 contributions to the workshop 'the physics of thermo-nuclear fusion and the ITER project': 1) the feasibility of magnetic confinement and the issue of heat recovery, 2) heating and current generation in tokamaks, 3) the physics of wall-plasma interaction, 4) recent results at JET, 5) inertial confinement and fast ignition, and 6) the technology of fusion machines based on magnetic confinement. This document presents the principles of thermo-nuclear fusion machines and gives a lot of technical information about JET, Tore-Supra and ITER.

ITER EDA Newsletter. V. 4, no. 3

International Nuclear Information System (INIS)

1995-03-01

This ITER EDA (Engineering Design Activities) Newsletter issue reports on (i) the ITER Magnet Technical Meeting held at the Naka Joint Work Site on February 7-10, 1995; (ii) the Second Technical Meeting on ITER Power Supply held on February 20-24, 1995, in St. Petersburg (Russian Federation); and (iii) a description by the Nuclear Data section of the IAEA (Vienna, Austria) on the availability and current status of the FENDL-1 Nuclear Data Libraries for fusion applications

ITER primary cryopump test facility

International Nuclear Information System (INIS)

Petersohn, N.; Mack, A.; Boissin, J.C.; Murdoc, D.

1998-01-01

A cryopump as ITER primary vacuum pump is being developed at FZK under the European fusion technology programme. The ITER vacuum system comprises of 16 cryopumps operating in a cyclic mode which fulfills the vacuum requirements in all ITER operation modes. Prior to the construction of a prototype cryopump, the concept is tested on a reduced scale model pump. To test the model pump, the TIMO facility is being built at FZK in which the model pump operation under ITER environmental conditions, except for tritium exposure, neutron irradiation and magnetic fields, can be simulated. The TIMO facility mainly consists of a test vessel for ITER divertor duct simulation, a 600 W refrigerator system supplying helium in the 5 K stage and a 30 kW helium supply system for the 80 K stage. The model pump test programme will be performed with regard to the pumping performance and cryogenic operation of the pump. The results of the model pump testing will lead to the design of the full scale ITER cryopump. (orig.)

The Physics Basis of ITER Confinement

International Nuclear Information System (INIS)

Wagner, F.

2009-01-01

ITER will be the first fusion reactor and the 50 year old dream of fusion scientists will become reality. The quality of magnetic confinement will decide about the success of ITER, directly in the form of the confinement time and indirectly because it decides about the plasma parameters and the fluxes, which cross the separatrix and have to be handled externally by technical means. This lecture portrays some of the basic principles which govern plasma confinement, uses dimensionless scaling to set the limits for the predictions for ITER, an approach which also shows the limitations of the predictions, and describes briefly the major characteristics and physics behind the H-mode--the preferred confinement regime of ITER.

Cochlear implant with a non-removable magnet: preliminary research at 3-T MRI.

Science.gov (United States)

Dubrulle, F; Sufana Iancu, A; Vincent, C; Tourrel, G; Ernst, O

2013-06-01

To perform preliminary tests in vitro and with healthy volunteers to determine the 3-T MRI compatibility of a cochlear implant with a non-removable magnet. In the in vitro phase, we tested six implants for temperature changes and internal malfunctioning. We measured the demagnetisation of 65 internal magnets with different tilt angles between the implant's magnetic field (bi) and the main magnetic field (b0). In the in vivo phase, we tested 28 operational implants attached to the scalps of volunteers with the head in three different positions. The study did not find significant temperature changes or electronic malfunction in the implants tested in vitro. We found considerable demagnetisation of the cochlear implant magnets in the in vitro and in vivo testing influenced by the position of the magnet in the main magnetic field. We found that if the bi/b0 angle is 90°, there is demagnetisation in almost 60 % of the cases. When the angle is around 90°, the risk of demagnetisation is low (6.6 %). The preliminary results on cochlear implants with non-removable magnets indicate the need to maintain the contraindication of passage through 3-T MRI. • Magnetic resonance imaging can affect cochlear implants and vice versa. • Demagnetisation of cochlear implant correlates with the angle between bi and b0. • The position of the head in the MRI influences the demagnetisation. • Three-Tesla MRI for cochlear implants is still contraindicated. • However some future solutions are discussed.

Status of ITER Cryodistribution and Cryoline project

Science.gov (United States)

Sarkar, B.; Vaghela, H.; Shah, N.; Bhattacharya, R.; Choukekar, K.; Patel, P.; Kapoor, H.; Srinivasa, M.; Chang, H. S.; Badgujar, S.; Monneret, E.

2017-02-01

The system of ITER Cryodistribution (CD) and Cryolines (CLs) is an integral interface between the Cryoplant systems and the superconducting (SC) magnets as well as Cryopumps (CPs). The project has progressed from the conceptual stage to the industrial stage. The subsystems are at various stages of design as defined by the project, namely, preliminary design, final design and formal reviews. Significant progresses have been made in the prototypes studies and design validations, such as the CL and cold circulators. While one of the prototype CL is already tested, the other one is in manufacturing phase. Performance test of two cold circulators have been completed. Design requirements are unique due the complexity arising from load specifications, layout constraints, regulatory compliance, operating conditions as well as several hundred interfaces. The present status of the project in terms of technical achievements, implications of the changes and the technical management as well as the risk assessment and its mitigation including path forward towards realization is described.

Conceptual design study of the K-DEMO magnet system

Energy Technology Data Exchange (ETDEWEB)

Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Oh, Sangjun; Park, Jong Sung; Lee, Chulhee; Im, Kihak; Kim, Hyung Chan; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Brown, Thomas; Kessel, Charles; Titus, Peter; Zhai, Yuhu [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

2015-10-15

Highlights: • Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. • Present a preliminary design of TF (toroidal field) magnet. • Present a preliminary design of CS (central solenoid) magnet. • Present a preliminary design of PF (toroidal field) magnet. - Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy. A major design philosophy for the initiated conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) is engineering feasibility. A two-staged development plan is envisaged. K-DEMO is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used, in its initial stage, as a component test facility. Then, in its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electricity generation on the order of 500 MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8 m and 2.1 m, respectively. In order to minimize wave deflection, a top-launch high frequency (>200 GHz) electron cyclotron current drive (ECCD) system will be the key system for the current profile control. For matching the high frequency ECCD, a high toroidal field (TF) is required and can be achieved by using high current density Nb{sub 3}Sn superconducting conductor. The peak magnetic field reaches to 16 T with the magnetic field at the plasma center above 7 T. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material.

Conceptual design study of the K-DEMO magnet system

International Nuclear Information System (INIS)

Kim, Keeman; Oh, Sangjun; Park, Jong Sung; Lee, Chulhee; Im, Kihak; Kim, Hyung Chan; Lee, Gyung-Su; Neilson, George; Brown, Thomas; Kessel, Charles; Titus, Peter; Zhai, Yuhu

2015-01-01

Highlights: • Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. • Present a preliminary design of TF (toroidal field) magnet. • Present a preliminary design of CS (central solenoid) magnet. • Present a preliminary design of PF (toroidal field) magnet. - Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy. A major design philosophy for the initiated conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) is engineering feasibility. A two-staged development plan is envisaged. K-DEMO is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used, in its initial stage, as a component test facility. Then, in its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electricity generation on the order of 500 MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8 m and 2.1 m, respectively. In order to minimize wave deflection, a top-launch high frequency (>200 GHz) electron cyclotron current drive (ECCD) system will be the key system for the current profile control. For matching the high frequency ECCD, a high toroidal field (TF) is required and can be achieved by using high current density Nb_3Sn superconducting conductor. The peak magnetic field reaches to 16 T with the magnetic field at the plasma center above 7 T. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material.

Preliminary piping layout and integration of European test blanket modules subsystems in ITER CVCS area

Energy Technology Data Exchange (ETDEWEB)

Tarallo, Andrea, E-mail: andrea.tarallo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Mozzillo, Rocco; Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Aiello, Antonio; Utili, Marco [ENEA UTIS, C.R. Brasimone, Bacino del Brasimone, I-40032 Camugnano, BO (Italy); Ricapito, Italo [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)

2015-04-15

Highlights: • The use of human modeling tools for piping design in view of maintenance is discussed. • A possible preliminary layout for TBM subsystems in CVCS area has been designed with CATIA. • A DHM-based method to quickly check for maintainability of piping systems is suggested. - Abstract: This paper explores a possible integration of some ancillary systems of helium-cooled lithium lead (HCLL) and helium-cooled pebble-bed (HCPB) test blanket modules in ITER CVCS area. Computer-aided design and ergonomics simulation tools have been fundamental not only to define suitable routes for pipes, but also to quickly check for maintainability of equipment and in-line components. In particular, accessibility of equipment and systems has been investigated from the very first stages of the design using digital human models. In some cases, the digital simulations have resulted in changes in the initial space reservations.

Optimization of Iter with Iter-89P scaling

International Nuclear Information System (INIS)

Johner, J.

1991-10-01

Ignition in the ITER baseline machine is studied in the frame of a 1/2-D model using the ITER-89P scaling of the energy confinement time. The required value of the enhancement factor f L with respect to the L-mode, allowing ignition with a total fusion power of 1100 MW, is found to be 1.9 at an optimum operating temperature of 11 keV. A sensitivity analysis shows that the critical f L =2 value can be exceeded with relatively small changes in the physical assumptions. It is concluded that the safety margin is not sufficient for this project. Optimization of a thermonuclear plasma in a tokamak is then performed with constraints of given maximum magnetic field B in the superconducting windings, given distance between the plasma and the maximum magnetic field point, imposed safety factor at the plasma edge, and given averaged neutron flux at the plasma surface. The minimum enhancement factor f L with respect to the L-mode, allowing ignition at a given value of the total fusion power P fus , is only a function of the torus aspect ratio A. Taking the ITER reference values for the above constraints, the required value of f L is practically independent of the aspect ratio but can be sensibly improved by increasing the total fusion power P fus . With P fus =1700 MW, a reasonable safety margin (f L ≅ 1.5) is obtained. Analytical expressions of the conditions resulting from the above optimization are also derived for an arbitrary monomial scaling of the energy confinement time, and shown to give excellent agreement with the numerical results

High magneticfield test of Bismuth Hall sensors for ITER steady state magnetic diagnostic

Czech Academy of Sciences Publication Activity Database

Ďuran, Ivan; Entler, Slavomír; Kohout, Michal; Kocan, M.; Vayakis, G.

2016-01-01

Roč. 87, č. 11 (2016), č. článku 11D446. ISSN 0034-6748. [Topical Conference on High-Temperature Plasma Diagnostics (HTPD2016) /21./. Madison, Wisconsin, 05.06.2016-09.06.2016] R&D Projects: GA MŠk LG14002 Institutional support: RVO:61389021 ; RVO:68378271 Keywords : Hall sensors * ITER * Hall effect * magnetic diagnostic Subject RIV: BL - Plasma and Gas Discharge Physics; BL - Plasma and Gas Discharge Physics (FZU-D) OBOR OECD: 2.11 Other engineering and technologies; 2.11 Other engineering and technologies (FZU-D) Impact factor: 1.515, year: 2016 http://scitation.aip.org/content/aip/journal/rsi/87/11/10.1063/1.4964435

Infrared laser diagnostics for ITER

International Nuclear Information System (INIS)

Hutchinson, D.P.; Richards, R.K.; Ma, C.H.

1995-01-01

Two infrared laser-based diagnostics are under development at ORNL for measurements on burning plasmas such as ITER. The primary effort is the development of a CO 2 laser Thomson scattering diagnostic for the measurement of the velocity distribution of confined fusion-product alpha particles. Key components of the system include a high-power, single-mode CO 2 pulsed laser, an efficient optics system for beam transport and a multichannel low-noise infrared heterodyne receiver. A successful proof-of-principle experiment has been performed on the Advanced Toroidal Facility (ATF) stellerator at ORNL utilizing scattering from electron plasma frequency satellites. The diagnostic system is currently being installed on Alcator C-Mod at MIT for measurements of the fast ion tail produced by ICRH heating. A second diagnostic under development at ORNL is an infrared polarimeter for Faraday rotation measurements in future fusion experiments. A preliminary feasibility study of a CO 2 laser tangential viewing polarimeter for measuring electron density profiles in ITER has been completed. For ITER plasma parameters and a polarimeter wavelength of 10.6 microm, a Faraday rotation of up to 26 degree is predicted. An electro-optic polarization modulation technique has been developed at ORNL. Laboratory tests of this polarimeter demonstrated a sensitivity of ≤ 0.01 degree. Because of the similarity in the expected Faraday rotation in ITER and Alcator C-Mod, a collaboration between ORNL and the MIT Plasma Fusion Center has been undertaken to test this polarimeter system on Alcator C-Mod. A 10.6 microm polarimeter for this measurement has been constructed and integrated into the existing C-Mod multichannel two-color interferometer. With present experimental parameters for C-Mod, the predicted Faraday rotation was on the order of 0.1 degree. Significant output signals were observed during preliminary tests. Further experiment and detailed analyses are under way

Development and test of the ITER conductor joints

Energy Technology Data Exchange (ETDEWEB)

Martovetsky, N., LLNL

1998-05-14

Joints for the ITER superconducting Central Solenoid should perform in rapidly varying magnetic field with low losses and low DC resistance. This paper describes the design of the ITER joint and presents its assembly process. Two joints were built and tested at the PTF facility at MIT. Test results are presented, losses in transverse and parallel field and the DC performance are discussed. The developed joint demonstrates sufficient margin for baseline ITER operating scenarios.

Analysis of the ITER cryoplant operational modes

International Nuclear Information System (INIS)

Henry, D.; Journeaux, J.Y.; Roussel, P.; Michel, F.; Poncet, J.M.; Girard, A.; Kalinin, V.; Chesny, P.

2007-01-01

In the framework of an EFDA task, CEA is carrying out an analysis of the various ITER cryoplant operational modes. According to the project integration document, ITER is designed to be operated 365 days per year in order to optimize the available time of the Tokamak. It is anticipated that operation will be performed in long periods separated by maintenance periods (e.g. 10 days continuous operation and 1 week break) with annual or bi-annual major shutdown periods of a few months for maintenance, further installation and commissioning. For this operation schedule, auxiliary subsystems like the cryoplant and the cryodistribution have to cope with different heat loads which depend on the different ITER operating states. The cryoplant consists of four identical 4.5 K refrigerators and two 80 K helium loops coupled with two LN2 modules. All of these cryogenic subsystems have to operate in parallel to remove the heat loads from the magnet, 80 K shields, cryopumps and other small users. After a brief recall of the main particularities of a cryogenic system operating in a Tokamak environment, the first part of this study is dedicated to the assessment of the main ITER operation states. A new design of refrigeration loop for the HTS current leads, the updated layout of the cryodistribution system and revised strategy for operations of the cryopumps have been taken into consideration. The relevant normal operating scenarios of the cryoplant are checked for the typical ITER operating states like plasma operation state, short term stand by, short term maintenance, or test and conditioning state. The second part of the paper is dedicated to the abnormal operating modes coming from the magnets and from those generated by the cryoplant itself. The occurrence of a fast discharge or a quench of the magnets generates large heat loads disturbances and produces exceptional high mass flow rates which have to be managed by the cryoplant, while a failure of a cryogenic component induces

Advanced scenarios for ITER operation

Energy Technology Data Exchange (ETDEWEB)

Sips, A.C.C. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

2004-07-01

In thermonuclear fusion research using magnetic confinement, the tokamak is the leading candidate for achieving conditions required for a reactor. An international experiment, ITER is proposed as the next essential and critical step on the path to demonstrating the scientific and technological feasibility of fusion energy. ITER is to produce and study plasmas dominated by self heating. This would give unique opportunities to explore, in reactor relevant conditions, the physics of {alpha}-particle heating, plasma turbulence and turbulent transport, stability limits to the plasma pressure and exhaust of power and particles. Important new results obtained in experiments, theory and modelling, enable an improved understanding of the physical processes occurring in tokamak plasmas and give enhanced confidence in ITER achieving its goals. In particular, progress has been made in research to raise the performance of tokamaks, aimed to extend the discharge pulse length towards steady-state operation (advanced scenarios). Standard tokamak discharges have a current density increasing monotonically towards the centre of the plasma. Advanced scenarios on the other hand use a modified current density profile. Different advanced scenarios range from (i) plasmas that sustain a central region with a flat current density profile (zero magnetic shear), capable of operating stationary at high plasma pressure, to (ii) discharges with an off axis maximum of the current density profile (reversed magnetic shear in the core), able to form internal transport barriers, to increase the confinement of the plasma. The physics of advanced tokamak discharges is described, together with an overview of recent results from different tokamak experiments. International collaboration between experiments aims to provide a better understanding, control and optimisation of these plasmas. The ability to explore advanced scenarios in ITER is very desirable, in order to verify the result obtained in

Meeting of the ITER CTA Project Board

International Nuclear Information System (INIS)

Vlasenkov, V.

2001-01-01

The meeting of the ITER CTA Project Board took place in Toronto, Canada on 7 November 2001, on the occasion of the first Negotiations Meeting. Twelve participants, representing PB members and experts from Canada, the EU, Japan, the RF and the international Team (IT), attended the meeting chaired by Acad. E. Velikhov. Discussions on the preliminary work programme for the CTA and organizational arrangements for the IT and PT took most of the time of the meeting. The Project Board approved the preliminary work programme as presented by the IT leader

Development of liquid metal type TBM technology for ITER

International Nuclear Information System (INIS)

Hong, Bong Guen; Kwak, J. G.; Kim, Y.

2008-03-01

The objectives of the ITER project for the construction and operation are to perform the test related to the neutronics, blanket module, tritium treatment technology, advanced plasma technology, and to test the heat extraction and tritium breeding in the test blanket for the fusion reactor. Other parties have been developing the Test Blanket Module (TBM) for testing in the ITER for these purposes. Through this project, we can secure the TBM design and related technology, which will be used as the core technology for the DEMO construction, our own fusion reactor development. In 1st year, the optimized design procedure was established with the existing tools, which have been used in nuclear reactor design, and the optimized HCML TBM design was obtained through iteration method according to the developed design procedure. He cooling system as a TBM auxiliary system was designed considering the final design of the KO HCML TBM such as coolant capacity and operation pressure. Layout for this system was prepared to be installed in the ITER TCWS vault. MHD effect of liquid Li breeder by magnetic flux in ITER such as much higher pressure drop was evaluated with CFD-ACE and it was concluded that the Li breeder should have a slow velocity to reduce this effect. Most results were arranged in the form of DDD including preliminary safety analysis report. In 2nd year, the optimized design procedure was complemented and updated. In performance analysis on thermal-hydraulic and thermo-mechanical one, full 3D meshes were generated and used in this analysis in order to obtain the more exact temperature, deformation, and stress solution. For liquid Li breeder system, design parameters were induced before the detailed design of the system and were used in the design of the liquid Li test loop. LOCA analysis, activation analysis in LOCA, EM analysis were performed as a preliminary safety analysis. In order to develop the manufacturing technology, Be+FMS and FMS to FMS joining conditions

Development of liquid metal type TBM technology for ITER

Energy Technology Data Exchange (ETDEWEB)

Hong, Bong Guen; Kwak, J. G.; Kim, Y. (and others)

2008-03-15

The objectives of the ITER project for the construction and operation are to perform the test related to the neutronics, blanket module, tritium treatment technology, advanced plasma technology, and to test the heat extraction and tritium breeding in the test blanket for the fusion reactor. Other parties have been developing the Test Blanket Module (TBM) for testing in the ITER for these purposes. Through this project, we can secure the TBM design and related technology, which will be used as the core technology for the DEMO construction, our own fusion reactor development. In 1st year, the optimized design procedure was established with the existing tools, which have been used in nuclear reactor design, and the optimized HCML TBM design was obtained through iteration method according to the developed design procedure. He cooling system as a TBM auxiliary system was designed considering the final design of the KO HCML TBM such as coolant capacity and operation pressure. Layout for this system was prepared to be installed in the ITER TCWS vault. MHD effect of liquid Li breeder by magnetic flux in ITER such as much higher pressure drop was evaluated with CFD-ACE and it was concluded that the Li breeder should have a slow velocity to reduce this effect. Most results were arranged in the form of DDD including preliminary safety analysis report. In 2nd year, the optimized design procedure was complemented and updated. In performance analysis on thermal-hydraulic and thermo-mechanical one, full 3D meshes were generated and used in this analysis in order to obtain the more exact temperature, deformation, and stress solution. For liquid Li breeder system, design parameters were induced before the detailed design of the system and were used in the design of the liquid Li test loop. LOCA analysis, activation analysis in LOCA, EM analysis were performed as a preliminary safety analysis. In order to develop the manufacturing technology, Be+FMS and FMS to FMS joining conditions

ITER, a major step toward nuclear fusion energy; ITER, une etape majeure vers l'energie de fusion

Energy Technology Data Exchange (ETDEWEB)

Ikeda, K.; Holtkamp, N.; Pick, M.; Gauche, F.; Garin, P.; Bigot, B.; Luciani, J.F.; Mougniot, J.C.; Watteau, J.P.; Saoutic, B.; Becoulet, A.; Libeyre, P.; Beaumont, B.; Simonin, A.; Giancarli, L.; Rosenvallon, S.; Gastaldi, O.; Marbach, G.; Boudot, C.; Ioki, K.; Mitchell, N.; Girard, J.Ph.; Giraud, B.; Lignini, F.; Giguet, E.; Bofusch, E.; Friconneau, J.P.; Di Pace, L.; Pampin, R.; Cook, I.; Maisonnier, D.; Campbell, D.; Hayward, J.; Li Puma, A.; Norajitra, P.; Sardain, P.; Tran, M.Q.; Ward, D.; Moslang, A.; Carre, F.; Serpantie, J.P

2007-01-15

This document gathers together a series of articles dedicated to ITER. They are organized into 5 parts. The first part describes the potential of fusion as a source of energy that will be able to face the challenge of a continuously increasing demand. After a reminder of the main fusion reactions and the conditions to obtain fusion, the second part focuses on the magnetic fusion based concepts with a special emphasis on the tokamak configuration. In the third part the main components of ITER are described: first the plasma facing components, then the vacuum vessel, the superconducting magnets and the heating systems. In the fourth part short papers concerning ITER safety, the maintenance through remote handling systems, the tritium breeding blanket, are given, along with a full article on the waste management. It is interesting to notice that the nuclear wastes will represent: -) between 1600 and 3800 tons of housekeeping and process wastes produced during the 20 years of operation of ITER (20% very low level waste, 75% low or medium activity with short life and 5% medium activity with long life), -) about 750 tons from component replacement during ITER active operation, and -) about 30000 tons from the decommissioning of ITER. The last part presents the European concepts for a power plant based on a fusion reactor. A basic design is given along with a state of the art of the research on the materials that will be used for the structures. It is highlighted that synergies between fission and fusion technologies exist in at least 4 areas: nuclear design code system, high temperature materials, safety approach, and in-service inspection, maintenance and dismantling. (A.C.)

The international magnetic fusion energy program

International Nuclear Information System (INIS)

Fowler, T.K.

1988-01-01

In May of 1988, the long tradition of international cooperation in magnetic fusion energy research culminated in the initiation of design work on the International Thermonuclear Experimental Reactor (ITER). If eventually constructed in the 1990s, ITER would be the world's first magnetic fusion reactor. This paper discusses the background events that led to ITER and the present status of the ITER activity. This paper presents a brief summary of the technical, political, and organizational activities that have led to the creation of the ITER design activity. The ITER activity is now the main focus of international cooperation in magnetic fusion research and one of the largest international cooperative efforts in all of science. 2 refs., 12 figs

Reactor-relevant quiescent H-mode operation using torque from non-axisymmetric, non-resonant magnetic fields

International Nuclear Information System (INIS)

Burrell, K. H.; Garofalo, A. M; Osborne, T. H.; Schaffer, M. J.; Snyder, P. B.; Solomon, W. M.; Park, J.-K.; Fenstermacher, M. E.

2012-01-01

Results from recent experiments demonstrate that quiescent H-mode (QH-mode) sustained by magnetic torque from non-axisymmetric magnetic fields is a promising operating mode for future burning plasmas. Using magnetic torque from n=3 fields to replace counter-I p torque from neutral beam injection (NBI), we have achieved long duration, counter-rotating QH-mode operation with NBI torque ranging from counter-I p to up to co-I p values of 1-1.3 Nm. This co-I p torque is 3 to 4 times the scaled torque that ITER will have. These experiments utilized an ITER-relevant lower single-null plasma shape and were done with ITER-relevant values of ν ped * and β N ped . These discharges exhibited confinement quality H 98y2 =1.3, in the range required for ITER. In preliminary experiments using n=3 fields only from a coil outside the toroidal coil, QH-mode plasmas with low q 95 =3.4 have reached fusion gain values of G=β N H 89 /q 95 2 =0.4, which is the desired value for ITER. Shots with the same coil configuration also operated with net zero NBI torque. The limits on G and co-I p torque have not yet been established for this coil configuration. QH-mode work to has made significant contact with theory. The importance of edge rotational shear is consistent with peeling-ballooning mode theory. Qualitative and quantitative agreements with the predicted neoclassical toroidal viscosity torque is seen.

R and D of the key components for ITER magnet supports

Energy Technology Data Exchange (ETDEWEB)

Lee, P.Y., E-mail: lipy@swip.ac.cn [Department of Material and Fusion Engineering, Southwestern Institute of Physics (SWIP), P.O. Box 432, Chengdu 610041 (China); Hou, B.L.; Pan, C.J. [Department of Material and Fusion Engineering, Southwestern Institute of Physics (SWIP), P.O. Box 432, Chengdu 610041 (China); Cheng, Z.Q.; Huai, K.W. [Stainless Steel Institute, Baoshan Iron and Steel Co., Ltd, 580 Changjiang Road, Baoshan District, Shanghai 200431 (China); Savary, F.; Fu, Y.K.; Gallix, R. [Tokamak Department, ITER Organization, Cadarache, 13108 St. Paul-Lez-Durance (France)

2010-12-15

In this article, we report the recent progress of the R and D work for ITER magnet supports, including 316LN raw material development, designing and manufacturing, development of gravity supports (GS) and poloidal field (PF) supports. 316LN stainless steel has been developed successfully with nitrogen content up to 0.188 mass%. These materials show very good mechanical properties at both room temperature and low temperature. The GS conceptual design using welded connections is replaced with pre-stressed bolted connections in order to reduce the risk of defects and manufacturing cost. The finite element model (FEM) analysis for this design shows that neither terrible stress concentrations nor large deformations would occur during either normal or abnormal operation. Attaching the thermal anchor to the flexible plate is made possible by a brazing connection and is carried out in the GS manufacturing. For PF supports, the wear resistance of the strut dowel of PF3-4 support improved after ion implantation.

R and D of the key components for ITER magnet supports

International Nuclear Information System (INIS)

Lee, P.Y.; Hou, B.L.; Pan, C.J.; Cheng, Z.Q.; Huai, K.W.; Savary, F.; Fu, Y.K.; Gallix, R.

2010-01-01

In this article, we report the recent progress of the R and D work for ITER magnet supports, including 316LN raw material development, designing and manufacturing, development of gravity supports (GS) and poloidal field (PF) supports. 316LN stainless steel has been developed successfully with nitrogen content up to 0.188 mass%. These materials show very good mechanical properties at both room temperature and low temperature. The GS conceptual design using welded connections is replaced with pre-stressed bolted connections in order to reduce the risk of defects and manufacturing cost. The finite element model (FEM) analysis for this design shows that neither terrible stress concentrations nor large deformations would occur during either normal or abnormal operation. Attaching the thermal anchor to the flexible plate is made possible by a brazing connection and is carried out in the GS manufacturing. For PF supports, the wear resistance of the strut dowel of PF3-4 support improved after ion implantation.

Meeting of the ITER CTA Project Board

International Nuclear Information System (INIS)

2001-01-01

Full text: A preparatory meeting of the Co-ordinated Technical Activities (CTA) Project Board took place in Vienna on 16 July 2001. The Board Members of Canada, EU, Japan, RF and of the CTA International Team participated in the Meeting, which was chaired by Acad. E. Velikhov. The major item on the Meeting Agenda was the discussion of the scope of the CTA. In this discussion the following comments were expressed: One of the prime objectives during the CTA is to develop technical specifications for procurement of critical items (magnets, vacuum vessel, and buildings). It was noted that the discussions with potential suppliers should confirm manufacturing processes in details in order to explore possible schedule reduction strategies. Safety analysis and licensing preparation should proceed on all proposed sites up to the preferred site designation, to ensure the overall implementation schedule is minimized and to resolve major technical issues needed for licensing. Several R and D issues remain to be further developed during the CTA. Special attention should be given by the Participants to two areas: Diagnostics; Heating and Current Drive Systems. Arrangements for continuation of the ITER Physics Expert Groups activities should be provided. To this end a new framework, called International Tokamak Physics Activity, is being planned. The Board encouraged the Participants' Representatives in the Co-ordinating committee of this activity to support the preparation for urgent Topical Group Meetings. The Board agreed that the Design Authority will be invested in the International Team and that proposals for site specific design changes should be agreed upon by the International Team Leader before being studied in detail. The Meeting agreed on some arrangements which will remain from the EDA, namely the ITER EDA Council Office in Moscow as Office of the PB Chair, and the ITER Office located at the IAEA in Vienna as agreed by the IAEA. The Board recommended that effective

Investigation of linearity of the ITER outer vessel steady-state magnetic field sensors at high temperature.

Czech Academy of Sciences Publication Activity Database

Entler, Slavomír; Ďuran, Ivan; Kocan, M.; Vayakis, G.

2017-01-01

Roč. 12, č. 7 (2017), č. článku C07007. ISSN 1748-0221. [European Conference on Plasma Diagnostics (ECPD2017)/2./. Bordeaux, 18.04.2017-21.04.2017] EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : Plasma diagnostics - probes * Detector alignment and calibration methods (lasers, sources, particle-beams) * ITER * Magnetic field Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.220, year: 2016 http://iopscience.iop.org/article/10.1088/1748-0221/12/07/C07007/pdf

Safety analysis results for cryostat ingress accidents in ITER

International Nuclear Information System (INIS)

Merrill, B.J.; Cadwallader, L.C.; Petti, D.A.

1996-01-01

Accidents involving the ingress of air or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits

FENIX [Fusion ENgineering International eXperimental]: A test facility for ITER [International Thermonuclear Experimental Reactor] and other new superconducting magnets

International Nuclear Information System (INIS)

Slack, D.S.; Patrick, R.E.; Miller, J.R.

1990-01-01

The Fusion ENgineering International eXperimental (FENIX) Test Facility which is nearing completion at Lawrence Livermore National Laboratory, is a 76-t set of superconducting magnets housed in a 4-m-diameter cryostat. It represents a significant step toward meeting the testing needs for the development of superconductors appropriate for large-scale magnet applications such as the International Thermonuclear Experimental Reactor (ITER). The magnet set is configured to allow radial access to the 0.4-m-diameter high-field region where maximum fields up to 14 T will be provided. The facility is fitted with a thermally isolated test well with a port to the high-field region that allows insertion and removal of test conductors without disturbing the cryogenic environment of the magnets. It is expected that the facility will be made available to magnet developers internationally, and this paper discusses its general design features, its construction, and its capabilities

ITER Central Solenoid Module Fabrication

Energy Technology Data Exchange (ETDEWEB)

Smith, John [General Atomics, San Diego, CA (United States)

2016-09-23

The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed, and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes: completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, and then stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power, along with, cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by five months of testing, which includes preliminary electrical testing followed by high current (48.5 kA) tests at 4.7K. The production of the seven modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first

MELCOR 1.8.2 Analyses in Support of ITER's RPrS

International Nuclear Information System (INIS)

Brad J Merrill

2008-01-01

The International Thermonuclear Experimental Reactor (ITER) Program is performing accident analyses for ITER's 'Rapport Preliminaire de Surete' (Report Preliminary on Safety - RPrS) with a modified version of the MELCOR 1.8.2 code. The RPrS is an ITER safety document required in the ITER licensing process to obtain a 'Decret Autorisation de Construction' (a Decree Authorizing Construction - DAC) for the ITER device. This report documents the accident analyses performed by the US with the MELCOR 1.8.2 code in support of the ITER RPrS effort. This work was funded through an ITER Task Agreement for MELCOR Quality Assurance and Safety Analyses. Under this agreement, the US was tasked with performing analyses for three accident scenarios in the ITER facility. Contained within the text of this report are discussions that identify the cause of these accidents, descriptions of how these accidents are likely to proceed, the method used to analyze the consequences of these accidents, and discussions of the transient thermal hydraulic and radiological release results for these accidents

A comparison between progressive extension method (PEM) and iterative method (IM) for magnetic field extrapolations in the solar atmosphere

Science.gov (United States)

Wu, S. T.; Sun, M. T.; Sakurai, Takashi

1990-01-01

This paper presents a comparison between two numerical methods for the extrapolation of nonlinear force-free magnetic fields, viz the Iterative Method (IM) and the Progressive Extension Method (PEM). The advantages and disadvantages of these two methods are summarized, and the accuracy and numerical instability are discussed. On the basis of this investigation, it is claimed that the two methods do resemble each other qualitatively.

Chapter 8: Plasma operation and control [Progress in the ITER Physics Basis (PIPB)

International Nuclear Information System (INIS)

Gribov, Y.; Humphreys, D.; Kajiwara, K.; Lazarus, E.A.; Lister, J.B.; Ozeki, T.; Portone, A.; Shimada, M.; Sips, A.C.C.; Wesley, J.C.

2007-01-01

The ITER plasma control system has the same functional scope as the control systems in present tokamaks. These are plasma operation scenario sequencing, plasma basic control (magnetic and kinetic), plasma advanced control (control of RWMs, NTMs, ELMs, error fields, etc) and plasma fast shutdown. This chapter considers only plasma initiation and plasma basic control. This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This assessment was done for the present ITER design (15 MA machine) at a more detailed level than it was done for the ITER design 1998 (21 MA machine) described in the ITER Physics Basis (1999 Nucl. Fusion 39 2577). The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, within specified assumptions on the plasma confinement and the impurity influx, ITER can produce plasma initiation in a low toroidal electric field (0.3 V m -1 ), if it is assisted by about 2 MW of ECRF heating. The plasma basic control includes control of the plasma current, position and shape-the plasma magnetic control, as well as control of other plasma global parameters or their profiles-the plasma performance control. The magnetic control is based on more reliable and simpler models of the control objects than those available at present for the plasma kinetic control. Moreover the real time diagnostics used for the magnetic control in many cases are more precise than those used for the kinetic control. Because of these reasons, the plasma magnetic control was developed for modern tokamaks and assessed for ITER better than the kinetic control. However, significant progress has been achieved in the plasma performance control during the last few years. Although the physics basis of plasma operation

Manufacturing and preliminary tests of a 12 T ''wind and react'' coil

International Nuclear Information System (INIS)

Corte, A. della; Pasotti, G.; Sacchetti, N.; Spadoni, M.; Oliva, A.B.; Penco, R.; Parodi, S.; Valle, N.; Specking, W.

1994-01-01

As already reported ENEA is engaged in the realization of a 12 T wind and react Nb 3 Sn coil, a subsize magnet designed to simulate many technological problems to be faced in NET-ITER magnets. EM-LMI and Ansaldo are the industrial partners in this project. A preliminary winding has been built and successfully tested. This winding has been cut in pieces and carefully inspected to be sure that the impregnation process after the heat treatment works well. No particular flaws have been detected. Then manufacturing of the 12 T magnet has been started and completed in about three months. Heat treatment, impregnation and electrical tests at 300 K have been successfully performed and the magnet is now ready for final tests. In order to obtain the most significant scientific and technological information from this magnet, the original test program (insertion of the coil in the SULTAN facility) has been modified according to a decision of the Fusion Technology Steering Committee (FTSC) of EURATOM. Details of the new test programs are given in the paper

Overview and status of ITER Cryostat manufacturing

Energy Technology Data Exchange (ETDEWEB)

Bhardwaj, Anil K., E-mail: anil.bhardwaj@iter-india.org [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Gupta, Girish; Prajapati, Rajnikant; Joshi, Vaibhav; Patel, Mitul; Bhavsar, Jagrut; More, Vipul; Jindal, Mukesh; Bhattacharya, Avik; Jogi, Gourav; Palaliya, Amit; Jha, Saroj; Pandey, Manish; Shukla, Dileep [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Iyer, Ganesh; Jadhav, Pandurang; Goyal, Dipesh; Desai, Anish [Larsen & Toubro Limited, Heavy Engineering, Hazira Manufacturing Complex, Gujarat (India); Sekachev, I.; Vitupier, Guillaume [ITER Organization, Route de Vinon sur Verdon – CS 90046, 13067 Saint Paul Lez Durance Cedex (France); and others

2016-11-01

Highlights: • Manufacturing status of one of the largest and the heaviest fully welded stainless steel vacuum chambers in the world (ITER Cryostat). • Overview of manufacturing stages and its segmentation. • Overview of manufacturing procedures and assembly and installation. - Abstract: One of ITER-India's commitments to the ITER Organization is procurement of the ITER Cryostat. It is a large vacuum vessel (∼29 m dia. and ∼29 m height), which is made up of 304/304 L dual marked stainless steel and has a total mass over 3500 t. The thickness of the vessel wall varies from 50 mm to 190 mm. It is one of the largest and the heaviest fully welded stainless steel vacuum chambers in the world which provides vacuum thermal insulation for the superconducting magnets operating at 4.5 K and for the thermal shield operating at 80 K. It also mechanically supports the magnet system along with the vacuum vessel (VV). The cryostat is designed and constructed according to ASME Section-VIII Division-2 with additional ITER Vacuum Handbook requirements and it is classified as protection important component (PIC-2). Manufacturing of cryostat segments is ongoing in India; sub-assembly of four major sections of the cryostat from the segments will be done at the ITER site in a temporary workshop building and the final assembly will be done in the pit of the tokamak building, the final location. The cryostat manufacturing contract has been awarded to Larsen and Toubro Limited in August 2012 after completion of design [4] and signing of Procurement Arrangement [1] with ITER Organization. Manufacturing of the cryostat was started in January 2014 after approval of the manufacturing drawings and procedures. The temporary workshop of 44 m × 110 m × 26 m in height has been completed in November 2014 at the ITER site with a 200 t crane installed. This paper gives an overview and the status of the cryostat manufacturing.

ITER EDA Newsletter. V. 3, no. 12

International Nuclear Information System (INIS)

1994-12-01

This ITER EDA (Engineering Design Activities) Newsletter issue reports on (i) the seventh Meeting of the Technical Advisory Committee (TAC-7) held at the Joint Work Site in Naka, Japan, 5-7 December 1994; (ii) the seventh Meeting of the ITER Management Advisory Committee (MAC-7) held at the Naka Joint Work Site, November 30 - December 1, 1994; (iii) the Magnet Technical Meeting, held at the Naka Joint Work Site on November 8-11, 1994

The international magnetic fusion energy program

Energy Technology Data Exchange (ETDEWEB)

Fowler, T.K.

1988-10-06

In May of 1988, the long tradition of international cooperation in magnetic fusion energy research culminated in the initiation of design work on the International Thermonuclear Experimental Reactor (ITER). If eventually constructed in the 1990s, ITER would be the world's first magnetic fusion reactor. This paper discusses the background events that led to ITER and the present status of the ITER activity. This paper presents a brief summary of the technical, political, and organizational activities that have led to the creation of the ITER design activity. The ITER activity is now the main focus of international cooperation in magnetic fusion research and one of the largest international cooperative efforts in all of science. 2 refs., 12 figs.

Status of ITER

International Nuclear Information System (INIS)

Aymar, R.

2002-01-01

At the end of engineering design activities (EDA) in July 2001, all the essential elements became available to make a decision on construction of ITER. A sufficiently detailed and integrated engineering design now exists for a generic site, has been assessed for feasibility, and costed, and essential physics and technology R and D has been carried out to underpin the design choices. Formal negotiations have now begun between the current participants--Canada, Euratom, Japan, and the Russian Federation--on a Joint Implementation Agreement for ITER which also establishes the legal entity to run ITER. These negotiations are supported on technical aspects by Coordinated Technical Activities (CTA), which maintain the integrity of the project, for the good of all participants, and concentrate on preparing for procurement by industry of the longest lead items, and for formal application for a construction license with the host country. This paper highlights the main features of the ITER design. With cryogenically-cooled magnets close to neutron-generating plasma, the design of shielding with adequate access via port plugs for auxiliaries such as heating and diagnostics, and of remote replacement and refurbishing systems for in-vessel components, are particularly interesting nuclear technology challenges. Making a safety case for ITER to satisfy potential regulators and to demonstrate, as far as possible at this stage, the environmental attractiveness of fusion as an energy source, is also important. The paper gives illustrative details on this work, and an update on the progress of technical preparations for construction, as well as the status of the above negotiations

Wall conditioning for ITER: Current experimental and modeling activities

Energy Technology Data Exchange (ETDEWEB)

Douai, D., E-mail: david.douai@cea.fr [CEA, IRFM, Association Euratom-CEA, 13108 St. Paul lez Durance (France); Kogut, D. [CEA, IRFM, Association Euratom-CEA, 13108 St. Paul lez Durance (France); Wauters, T. [LPP-ERM/KMS, Association Belgian State, 1000 Brussels (Belgium); Brezinsek, S. [FZJ, Institut für Energie- und Klimaforschung Plasmaphysik, 52441 Jülich (Germany); Hagelaar, G.J.M. [Laboratoire Plasma et Conversion d’Energie, UMR5213, Toulouse (France); Hong, S.H. [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Lomas, P.J. [CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Lyssoivan, A. [LPP-ERM/KMS, Association Belgian State, 1000 Brussels (Belgium); Nunes, I. [Associação EURATOM-IST, Instituto de Plasmas e Fusão Nuclear, 1049-001 Lisboa (Portugal); Pitts, R.A. [ITER International Organization, F-13067 St. Paul lez Durance (France); Rohde, V. [Max-Planck-Institut für Plasmaphysik, 85748 Garching (Germany); Vries, P.C. de [ITER International Organization, F-13067 St. Paul lez Durance (France)

2015-08-15

Wall conditioning will be required in ITER to control fuel and impurity recycling, as well as tritium (T) inventory. Analysis of conditioning cycle on the JET, with its ITER-Like Wall is presented, evidencing reduced need for wall cleaning in ITER compared to JET–CFC. Using a novel 2D multi-fluid model, current density during Glow Discharge Conditioning (GDC) on the in-vessel plasma-facing components (PFC) of ITER is predicted to approach the simple expectation of total anode current divided by wall surface area. Baking of the divertor to 350 °C should desorb the majority of the co-deposited T. ITER foresees the use of low temperature plasma based techniques compatible with the permanent toroidal magnetic field, such as Ion (ICWC) or Electron Cyclotron Wall Conditioning (ECWC), for tritium removal between ITER plasma pulses. Extrapolation of JET ICWC results to ITER indicates removal comparable to estimated T-retention in nominal ITER D:T shots, whereas GDC may be unattractive for that purpose.

ITER status, design and material objectives

International Nuclear Information System (INIS)

Aymar, R.

2002-01-01

During the ITER Engineering Design Activities (EDA), completed in July 2001, the Joint Central Team and Home Teams developed a robust design of ITER, summarised in this paper, with parameters which fully meet the required scientific and technological objectives, construction costs and safety requirements, with appropriate margins. The design is backed by R and D to qualify the technology, including materials R and D. Materials for ITER components have been selected largely because of their availability and well-established manufacturing technologies, taking account of the low fluence experienced during neutron irradiation, and the experimental nature of the device. Nevertheless, for specific needs relevant to a future fusion reactor, improved materials, in particular for magnet structures, in-vessel components, and joints between the different materials needed for plasma facing components, have been successfully developed. Now, with the technical readiness to decide on ITER construction, negotiations, supported by coordinated technical activities of an international team and teams from participant countries, are underway on joint construction of ITER with a view to the signature/ratification of an agreement in 2003

Industrial opportunities on the International Thermonuclear Experimental Reactor (ITER) project

International Nuclear Information System (INIS)

Ellis, W.R.

1996-01-01

Industry has been a long-term contributor to the magnetic fusion program, playing a variety of important roles over the years. Manufacturing firms, engineering-construction companies, and the electric utility industry should all be regarded as legitimate stakeholders in the fusion energy program. In a program focused primarily on energy production, industry's future roles should follow in a natural way, leading to the commercialization of the technology. In a program focused primarily on science and technology, industry's roles, in the near term, should be, in addition to operating existing research facilities, largely devoted to providing industrial support to the International Thermonuclear Experimental Reactor (ITER) Project. Industrial opportunities on the ITER Project will be guided by the amount of funding available to magnetic fusion generally, since ITER is funded as a component of that program. The ITER Project can conveniently be discussed in terms of its phases, namely, the present Engineering Design Activities (EDA) phase, and the future (as yet not approved) construction phase. 2 refs., 3 tabs

Analysis of active and passive magnetic field reduction systems (MFRS) of the ITER NBI

International Nuclear Information System (INIS)

Roccella, M.; Lucca, F.; Roccella, R.; Pizzuto, A.; Ramogida, G.; Portone, A.; Tanga, A.; Formisano, A.; Martone, R.

2007-01-01

In ITER two heating (HNBI) and one diagnostic neutral beam injectors (DNBI) are foreseen. Inside these components there are very stringent limits on the magnetic field (the flux density must be below some G along the ion path and below 20 G in the neutralizing regions). To achieve these performances in an environment with high stray field due to the plasma and the poloidal field coils (PFC), both passive and active shielding systems have been foreseen. The present design of the magnetic field reduction systems (MFRS) is made of seven active coils and of a box surrounding the NBI region, consisting of ferromagnetic plates. The electromagnetic analyses of the effectiveness of these shields have been performed by a 3D FEM model using ANSYS code for the HNBI. The ANSYS models of the ferromagnetic box and of the active coils are fully parametric, thus any size change of the ferromagnetic box and coils (linear dimension or thickness) preserving the overall box shape could be easily reproduced by simply changing some parameter in the model

Analysis of active and passive magnetic field reduction systems (MFRS) of the ITER NBI

Energy Technology Data Exchange (ETDEWEB)

Roccella, M. [L.T. Calcoli S.a.S., Piazza Prinetti 26/B, Merate (Lecco) (Italy)], E-mail: roccella@ltcalcoli.it; Lucca, F.; Roccella, R. [L.T. Calcoli S.a.S., Piazza Prinetti 26/B, Merate (Lecco) (Italy); Pizzuto, A.; Ramogida, G. [Associazione EURATOM sulla Fusione - ENEA Frascati (Italy); Portone, A.; Tanga, A. [ITER EFDA (Italy); Formisano, A.; Martone, R. [CREATE Napoli (Italy)

2007-10-15

In ITER two heating (HNBI) and one diagnostic neutral beam injectors (DNBI) are foreseen. Inside these components there are very stringent limits on the magnetic field (the flux density must be below some G along the ion path and below 20 G in the neutralizing regions). To achieve these performances in an environment with high stray field due to the plasma and the poloidal field coils (PFC), both passive and active shielding systems have been foreseen. The present design of the magnetic field reduction systems (MFRS) is made of seven active coils and of a box surrounding the NBI region, consisting of ferromagnetic plates. The electromagnetic analyses of the effectiveness of these shields have been performed by a 3D FEM model using ANSYS code for the HNBI. The ANSYS models of the ferromagnetic box and of the active coils are fully parametric, thus any size change of the ferromagnetic box and coils (linear dimension or thickness) preserving the overall box shape could be easily reproduced by simply changing some parameter in the model.

An iterative homogenization technique that preserves assembly core exchanges

International Nuclear Information System (INIS)

Mondot, Ph.; Sanchez, R.

2003-01-01

A new interactive homogenization procedure for reactor core calculations is proposed that requires iterative transport assembly and diffusion core calculations. At each iteration the transport solution of every assembly type is used to produce homogenized cross sections for the core calculation. The converged solution gives assembly fine multigroup transport fluxes that preserve macro-group assembly exchanges in the core. This homogenization avoids the periodic lattice-leakage model approximation and gives detailed assembly transport fluxes without need of an approximated flux reconstruction. Preliminary results are given for a one-dimensional core model. (authors)

Heating and current drive requirements towards steady state operation in ITER

Energy Technology Data Exchange (ETDEWEB)

Poli, F. M.; Kessel, C. E.; Gorelenkova, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Bonoli, P. T. [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Batchelor, D. B. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Harvey, B.; Petrov, Y. [CompX, Box 2672, Del Mar, CA 92014 (United States)

2014-02-12

Steady state scenarios envisaged for ITER aim at optimizing the bootstrap current, while maintaining sufficient confinement and stability to provide the necessary fusion yield. Non-inductive scenarios will need to operate with Internal Transport Barriers (ITBs) in order to reach adequate fusion gain at typical currents of 9 MA. However, the large pressure gradients associated with ITBs in regions of weak or negative magnetic shear can be conducive to ideal MHD instabilities, reducing the no-wall limit. The E × B flow shear from toroidal plasma rotation is expected to be low in ITER, with a major role in the ITB dynamics being played by magnetic geometry. Combinations of H/CD sources that maintain weakly reversed magnetic shear profiles throughout the discharge are the focus of this work. Time-dependent transport simulations indicate that, with a trade-off of the EC equatorial and upper launcher, the formation and sustainment of quasi-steady state ITBs could be demonstrated in ITER with the baseline heating configuration. However, with proper constraints from peeling-ballooning theory on the pedestal width and height, the fusion gain and the maximum non-inductive current are below the ITER target. Upgrades of the heating and current drive system in ITER, like the use of Lower Hybrid current drive, could overcome these limitations, sustaining higher non-inductive current and confinement, more expanded ITBs which are ideal MHD stable.

Heating and current drive requirements towards steady state operation in ITER

Science.gov (United States)

Poli, F. M.; Bonoli, P. T.; Kessel, C. E.; Batchelor, D. B.; Gorelenkova, M.; Harvey, B.; Petrov, Y.

2014-02-01

Steady state scenarios envisaged for ITER aim at optimizing the bootstrap current, while maintaining sufficient confinement and stability to provide the necessary fusion yield. Non-inductive scenarios will need to operate with Internal Transport Barriers (ITBs) in order to reach adequate fusion gain at typical currents of 9 MA. However, the large pressure gradients associated with ITBs in regions of weak or negative magnetic shear can be conducive to ideal MHD instabilities, reducing the no-wall limit. The E × B flow shear from toroidal plasma rotation is expected to be low in ITER, with a major role in the ITB dynamics being played by magnetic geometry. Combinations of H/CD sources that maintain weakly reversed magnetic shear profiles throughout the discharge are the focus of this work. Time-dependent transport simulations indicate that, with a trade-off of the EC equatorial and upper launcher, the formation and sustainment of quasi-steady state ITBs could be demonstrated in ITER with the baseline heating configuration. However, with proper constraints from peeling-ballooning theory on the pedestal width and height, the fusion gain and the maximum non-inductive current are below the ITER target. Upgrades of the heating and current drive system in ITER, like the use of Lower Hybrid current drive, could overcome these limitations, sustaining higher non-inductive current and confinement, more expanded ITBs which are ideal MHD stable.

The ITER poloidal field system: control and power supplies

International Nuclear Information System (INIS)

Mondino, P.L.; Benfatto, I.; Gribov, Y.; Matsukawa, M.; Odajima, K.; Portone, A.; Roshal, A.; Bareyt, B.; Bertolini, E.; Bottereau, J.M.; Huart, M.; Maschio, A.; Bulgakov, S.; Kuchinski, V.

1995-01-01

The paper reports the preliminary scenario of the ITER Poloidal Field (PF) system operation, the method used to evaluate the installed power, the basic structure of the circuits and finally the concepts of the preliminary design of control and power supply. The superconducting coils are energized from the HV Grid with conventional AC/DC converters. R and D is required for circuit breakers, make switches and resistors, the basic components of both the switching networks and the discharge circuits. (orig.)

ITER operating limit definition criteria

International Nuclear Information System (INIS)

Ciattaglia, S.; Barabaschi, P.; Carretero, J.A.; Chiocchio, S.; Hureau, D.; Girard, J.Ph.; Gordon, C.; Portone, A.; Rodrigo, L. Rodriguez; Roldan, C.; Saibene, G.; Uzan-Elbez, J.

2009-01-01

The operating limits and conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which, together, define the domain of the safe operation of ITER in all foreseen ITER states (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components that are important for personnel or public safety (safety important class, SIC) are identified considering their functional importance in the overall plant safety analysis. SIC classification has to be presented already in the preliminary safety analysis report and approved by the licensing authority before manufacturing and construction. OLCs comprise the safety limits that, if exceeded, could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems, and the operational limits on equipment which warn against or stop a functional deviation from a planned operational status that could challenge equipment and functions. Some operational conditions, e.g. in-Vacuum Vessel (VV) radioactive inventories, will be controlled through procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants, is considered to the maximum possible extent. This paper presents the guidelines for the development of the ITER OLCs with particular reference to safety limits.

Three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

Czech Academy of Sciences Publication Activity Database

Schmitz, O.; Becoulet, M.; Cahyna, Pavel; Evans, T.E.; Feng, Y.; Frerichs, H.; Loarte, A.; Pitts, R.A.; Reiser, D.; Fenstermacher, M.E.; Harting, D.; Kirschner, A.; Kukushkin, A.; Lunt, T.; Saibene, G.; Reiter, D.; Samm, U.; Wiesen, S.

2016-01-01

Roč. 56, č. 6 (2016), č. článku 066008. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : resonant magnetic perturbations * plasma edge physics * 3D modeling * neutral particle physics * ITER * divertor heat and particle loads * ELM control Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/6/066008/meta

An Asdex-type divertor for ITER

International Nuclear Information System (INIS)

Fowler, T.K.

1989-01-01

An Asdex-type local divertor is proposed for ITER consisting of a copper poloidal field coil adjacent to the plasma. Estimates indicate that the power consumption is acceptable. Advantages would be a much reduced heat load not very sensitive to magnetic perturbations. A disadvantage is the finite lifetime under neutron bombardment that would require periodic replacement of the divertor coils in a reactor, but probably not in ITER because of its limited fluence. Another disadvantage would be poorer blanket coverage unless the divertor coil itself incorporates breeding material. 3 figs

A Novel Compressed Sensing Method for Magnetic Resonance Imaging: Exponential Wavelet Iterative Shrinkage-Thresholding Algorithm with Random Shift

Directory of Open Access Journals (Sweden)

Yudong Zhang

2016-01-01

Full Text Available Aim. It can help improve the hospital throughput to accelerate magnetic resonance imaging (MRI scanning. Patients will benefit from less waiting time. Task. In the last decade, various rapid MRI techniques on the basis of compressed sensing (CS were proposed. However, both computation time and reconstruction quality of traditional CS-MRI did not meet the requirement of clinical use. Method. In this study, a novel method was proposed with the name of exponential wavelet iterative shrinkage-thresholding algorithm with random shift (abbreviated as EWISTARS. It is composed of three successful components: (i exponential wavelet transform, (ii iterative shrinkage-thresholding algorithm, and (iii random shift. Results. Experimental results validated that, compared to state-of-the-art approaches, EWISTARS obtained the least mean absolute error, the least mean-squared error, and the highest peak signal-to-noise ratio. Conclusion. EWISTARS is superior to state-of-the-art approaches.

A Novel Compressed Sensing Method for Magnetic Resonance Imaging: Exponential Wavelet Iterative Shrinkage-Thresholding Algorithm with Random Shift

Science.gov (United States)

Zhang, Yudong; Yang, Jiquan; Yang, Jianfei; Liu, Aijun; Sun, Ping

2016-01-01

Aim. It can help improve the hospital throughput to accelerate magnetic resonance imaging (MRI) scanning. Patients will benefit from less waiting time. Task. In the last decade, various rapid MRI techniques on the basis of compressed sensing (CS) were proposed. However, both computation time and reconstruction quality of traditional CS-MRI did not meet the requirement of clinical use. Method. In this study, a novel method was proposed with the name of exponential wavelet iterative shrinkage-thresholding algorithm with random shift (abbreviated as EWISTARS). It is composed of three successful components: (i) exponential wavelet transform, (ii) iterative shrinkage-thresholding algorithm, and (iii) random shift. Results. Experimental results validated that, compared to state-of-the-art approaches, EWISTARS obtained the least mean absolute error, the least mean-squared error, and the highest peak signal-to-noise ratio. Conclusion. EWISTARS is superior to state-of-the-art approaches. PMID:27066068

Controlled Nuclear Fusion by Magnetic Confinement and ITER

CERN Document Server

CERN. Geneva. Audiovisual Unit; Alvarez-Gaumé, Luís

2005-01-01

For may years harnessing fusion energy was considered the final solution to the world's energy crisis. ITER is the last step in the elusive quest. This presentation will provide in its various acientific, technological and political aspects.

Nuclear analysis for ITER

International Nuclear Information System (INIS)

Santoro, R.T.; Iida, H.; Khripunov, V.; Petrizzi, L.; Sato, S.; Sawan, M.; Shatalov, G.; Schipakin, O.

2001-01-01

This paper summarizes the main results of nuclear analysis calculations performed during the International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA). Major efforts were devoted to fulfilling the General Design Requirements to minimize the nuclear heating rate in the superconducting magnets and ensuring that radiation conditions at the cryostat are suitable for hands-on-maintenance after reactor shut-down. (author)

ITER Operating Limits and Conditions

International Nuclear Information System (INIS)

Ciattaglia, S.; Barabaschi, P.; Carretero, J.A.

2006-01-01

The Operating Limits and Conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which together define the domain of the safe operation of ITER in all foreseen ITER status (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components important for personnel or public safety (Safety Important Class, SIC) are identified from the overall plant safety analysis on functional importance to safety of the components. SIC classification has to be presented already inside the preliminary safety analysis report and approved by the licensing safety authority before the relevant construction. OLCs comprise the safety limits, i.e. that if exceeded could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems and the operational limits on equipment which warn from or stop a functional departure from a planned operational status that could challenge equipment and functions. The safety limits have to indicate clearly states that leave the nominal safety state of ITER; they are derived from the safety analysis of ITER. OLCs can represent in some cases few parameters grouping together. Some operational conditions, e.g. inventories, will be controlled through no real time measurements and procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants is considered at the maximum possible extent. This paper presents the guidelines to develop the ITER OLCs with particular reference to safety limits. A few examples are reported as well as open issues on some OLCs control and measurement and the relevant R-and-D planned to solve the issues. (author)

Application of the perturbation iteration method to boundary layer type problems.

Science.gov (United States)

Pakdemirli, Mehmet

2016-01-01

The recently developed perturbation iteration method is applied to boundary layer type singular problems for the first time. As a preliminary work on the topic, the simplest algorithm of PIA(1,1) is employed in the calculations. Linear and nonlinear problems are solved to outline the basic ideas of the new solution technique. The inner and outer solutions are determined with the iteration algorithm and matched to construct a composite expansion valid within all parts of the domain. The solutions are contrasted with the available exact or numerical solutions. It is shown that the perturbation-iteration algorithm can be effectively used for solving boundary layer type problems.

Manufacture and test of prototype water pipe chase barrier in ITER Magnet Feeder system

Energy Technology Data Exchange (ETDEWEB)

Lu, Kun, E-mail: lukun@ipp.ac.cn [Institute of Plasma Physics, Shushan Hu Road 350, Hefei, Anhui (China); Wen, Xinjie; Liu, Chen; Song, Yuntao [Institute of Plasma Physics, Shushan Hu Road 350, Hefei, Anhui (China); Niu, Erwu [ITER China, 15B Fuxing Road, Beijing 100862 (China); Gung, Chenyu; Su, Man [ITER Organization, Route de Vinon-sur-Verdon – CS 90046, 13067 St Paul-lez-Durance Cedex (France)

2016-11-01

The Magnet Feeder system in the International Thermonuclear Experimental Reactor (ITER) deploys electrical currents and supercritical helium to the superconducting magnets and the magnet diagnostic signals to the operators. In the current design, the feeders located in the upper L3 level of the Tokamak gallery penetrate the Tokamak coolant water system vault, the biological shield and the cryostat. As a secondary confinement to contain the activated coolant water in the vault in the case of water pipe burst accident, a water barrier is welded between the penetration in the water pipe chase outer wall and the mid-plane of the vacuum jacket of the Feeder Coil Terminal Box (CTB). A thin-wall stainless steel diaphragm with an omega shape profile is welded around the CTB as the water barrier to endure 2 bar hydraulic pressure. In addition, the barrier is designed as a flexible compensator to withstand a maximum of 15 mm of axial displacement of the CTB in case of helium leak accident without failure. This paper presents the detail configuration, the manufacturing and assembly processes of the water barrier. Test results of the prototype water barrier under simulated accident conditions are also reported. Successful qualification of the design and manufacturing process of the water barrier lays a good foundation for the series production of this subsystem.

Pseudoinverse preconditioners and iterative methods for large dense linear least-squares problems

Directory of Open Access Journals (Sweden)

Oskar Cahueñas

2013-05-01

Full Text Available We address the issue of approximating the pseudoinverse of the coefficient matrix for dynamically building preconditioning strategies for the numerical solution of large dense linear least-squares problems. The new preconditioning strategies are embedded into simple and well-known iterative schemes that avoid the use of the, usually ill-conditioned, normal equations. We analyze a scheme to approximate the pseudoinverse, based on Schulz iterative method, and also different iterative schemes, based on extensions of Richardson's method, and the conjugate gradient method, that are suitable for preconditioning strategies. We present preliminary numerical results to illustrate the advantages of the proposed schemes.

Outlier Detection in Regression Using an Iterated One-Step Approximation to the Huber-Skip Estimator

DEFF Research Database (Denmark)

Johansen, Søren; Nielsen, Bent

2013-01-01

In regression we can delete outliers based upon a preliminary estimator and reestimate the parameters by least squares based upon the retained observations. We study the properties of an iteratively defined sequence of estimators based on this idea. We relate the sequence to the Huber-skip estima......In regression we can delete outliers based upon a preliminary estimator and reestimate the parameters by least squares based upon the retained observations. We study the properties of an iteratively defined sequence of estimators based on this idea. We relate the sequence to the Huber...... that the normalized estimation errors are tight and are close to a linear function of the kernel, thus providing a stochastic expansion of the estimators, which is the same as for the Huber-skip. This implies that the iterated estimator is a close approximation of the Huber-skip...

ITER: the Sun rises over nuclear fusion with West

International Nuclear Information System (INIS)

Sacco, Laurent

2013-01-01

The ITER project is considered as a critical step on the way to commercial production of electricity by a thermonuclear reactor based on controlled fusion. This project notably requires the development of a divertor which is the objective of the West project which will use the famous Cadarache superconductive magnet reactor, Tore Supra. After having outlined the future lack of fossil energies at the world scale, presented the operation principles of tokamaks and recalled some results obtained in their development, this article justifies the use of superconductive magnets. It presents the ITER project as a step in the production of thermonuclear electricity. ITER will be in fact a proof that such plants can be realised, and it should be followed by Demo, a demonstration power plant, by 2050. The article presents the West project, a test bench for ITER, which introduced modifications in the Tore Supra reactor to create conditions almost similar to that existing at the surface of the Sun. It notably comprises a divertor made of tungsten for the fusion with tritium. It finally outlines that the fusion will be a hot one, not a cold one

ITER, a major step toward nuclear fusion energy

International Nuclear Information System (INIS)

Ikeda, K.; Holtkamp, N.; Pick, M.; Gauche, F.; Garin, P.; Bigot, B.; Luciani, J.F.; Mougniot, J.C.; Watteau, J.P.; Saoutic, B.; Becoulet, A.; Libeyre, P.; Beaumont, B.; Simonin, A.; Giancarli, L.; Rosenvallon, S.; Gastaldi, O.; Marbach, G.; Boudot, C.; Ioki, K.; Mitchell, N.; Girard, J.Ph.; Giraud, B.; Lignini, F.; Giguet, E.; Bofusch, E.; Friconneau, J.P.; Di Pace, L.; Pampin, R.; Cook, I.; Maisonnier, D.; Campbell, D.; Hayward, J.; Li Puma, A.; Norajitra, P.; Sardain, P.; Tran, M.Q.; Ward, D.; Moslang, A.; Carre, F.; Serpantie, J.P.

2007-01-01

This document gathers together a series of articles dedicated to ITER. They are organized into 5 parts. The first part describes the potential of fusion as a source of energy that will be able to face the challenge of a continuously increasing demand. After a reminder of the main fusion reactions and the conditions to obtain fusion, the second part focuses on the magnetic fusion based concepts with a special emphasis on the tokamak configuration. In the third part the main components of ITER are described: first the plasma facing components, then the vacuum vessel, the superconducting magnets and the heating systems. In the fourth part short papers concerning ITER safety, the maintenance through remote handling systems, the tritium breeding blanket, are given, along with a full article on the waste management. It is interesting to notice that the nuclear wastes will represent: -) between 1600 and 3800 tons of housekeeping and process wastes produced during the 20 years of operation of ITER (20% very low level waste, 75% low or medium activity with short life and 5% medium activity with long life), -) about 750 tons from component replacement during ITER active operation, and -) about 30000 tons from the decommissioning of ITER. The last part presents the European concepts for a power plant based on a fusion reactor. A basic design is given along with a state of the art of the research on the materials that will be used for the structures. It is highlighted that synergies between fission and fusion technologies exist in at least 4 areas: nuclear design code system, high temperature materials, safety approach, and in-service inspection, maintenance and dismantling. (A.C.)

Rotation and neoclassical ripple transport in ITER

Science.gov (United States)

Paul, E. J.; Landreman, M.; Poli, F. M.; Spong, D. A.; Smith, H. M.; Dorland, W.

2017-11-01

Neoclassical transport in the presence of non-axisymmetric magnetic fields causes a toroidal torque known as neoclassical toroidal viscosity (NTV). The toroidal symmetry of ITER will be broken by the finite number of toroidal field coils and by test blanket modules (TBMs). The addition of ferritic inserts (FIs) will decrease the magnitude of the toroidal field ripple. 3D magnetic equilibria in the presence of toroidal field ripple and ferromagnetic structures are calculated for an ITER steady-state scenario using the variational moments equilibrium code (VMEC). Neoclassical transport quantities in the presence of these error fields are calculated using the stellarator Fokker-Planck iterative neoclassical conservative solver (SFINCS). These calculations fully account for E r , flux surface shaping, multiple species, magnitude of ripple, and collisionality rather than applying approximate analytic NTV formulae. As NTV is a complicated nonlinear function of E r , we study its behavior over a plausible range of E r . We estimate the toroidal flow, and hence E r , using a semi-analytic turbulent intrinsic rotation model and NUBEAM calculations of neutral beam torque. The NTV from the \\vert{n}\\vert = 18 ripple dominates that from lower n perturbations of the TBMs. With the inclusion of FIs, the magnitude of NTV torque is reduced by about 75% near the edge. We present comparisons of several models of tangential magnetic drifts, finding appreciable differences only for superbanana-plateau transport at small E r . We find the scaling of calculated NTV torque with ripple magnitude to indicate that ripple-trapping may be a significant mechanism for NTV in ITER. The computed NTV torque without ferritic components is comparable in magnitude to the NBI and intrinsic turbulent torques and will likely damp rotation, but the NTV torque is significantly reduced by the planned ferritic inserts.

An integrated model for the assessment of unmitigated fault events in ITER's superconducting magnets

Energy Technology Data Exchange (ETDEWEB)

McIntosh, S., E-mail: simon.mcintosh@ccfe.ac.uk [Culham Centre for Fusion Energy, Culham Science Center, Abingdon OX14 3DB, Oxfordshire (United Kingdom); Holmes, A. [Marcham Scientific Ltd., Sarum House, 10 Salisbury Rd., Hungerford RG17 0LH, Berkshire (United Kingdom); Cave-Ayland, K.; Ash, A.; Domptail, F.; Zheng, S.; Surrey, E.; Taylor, N. [Culham Centre for Fusion Energy, Culham Science Center, Abingdon OX14 3DB, Oxfordshire (United Kingdom); Hamada, K.; Mitchell, N. [ITER Organization, Magnet Division, St Paul Lez Durance Cedex (France)

2016-11-01

A large amount of energy is stored in ITER superconducting magnet system. Faults which initiate a discharge are typically mitigated to quickly transfer away the stored magnetic energy for dissipation through a bank of resistors. In an extreme unlikely occurrence, an unmitigated fault event represents a potentially severe discharge of energy into the coils and the surrounding structure. A new simulation tool has been developed for the detailed study of these unmitigated fault events. The tool integrates: the propagation of multiple quench fronts initiated by an initial fault or by subsequent coil heating; the 3D convection and conduction of heat through the magnet structure; the 3D conduction of current and Ohmic heating both along the conductor and via alternate pathways generated by arcing or material melt. Arcs linking broken sections of conductor or separate turns are simulated with a new unconstrained arc model to balance electrical current paths and heat generation within the arc column in the multi-physics model. The influence under the high Lorenz forces present is taken into account. Simulation results for an unmitigated fault in a poloidal field coil are presented.

The ITER poloidal field configuration and operation scenario

International Nuclear Information System (INIS)

Gribov, Y.; Portone, A.; Mondino, P.L.

1995-01-01

The ITER Poloidal Field (PF) system must satisfy the following requirements. (1) ITER must have a well-controlled, single null divertor magnetic configuration with nominal plasma current 21MA and moderate plasma elongation k95 < 1.65. (2) For a variety of plasma scenarios the ITER PF system must provide: inductive breakdown and start-up in an expanding-aperture limiter configuration near the outboard first wall; an inductive current ramp-up to the nominal plasma current with a reasonable assumption of resistive loss during current ramp-up; a pulse length of 1,000s for ignition and inductively-sustained burn at nominal plasma current; plasma shutdown (following fusion power termination) in a similar contracting-aperture limiter configuration. The present design of the PF system can satisfy the ITER requirements within specified limitations

Gamma irradiation testing of prototype ITER in-vessel magnetic pick-up coils

International Nuclear Information System (INIS)

Vermeeren, Ludo; Leysen, Willem

2013-01-01

Highlights: ► We tested five prototype ITER in-vessel coils up to a gamma dose of 72 MGy. ► Before and after irradiation thermal tests were also performed from 30 °C till 130 °C. ► The continuity resistances and the insulation resistances were continuously monitored. ► The observed behavior of all coils was satisfactory in all conditions. ► For the further design the mechanical robustness should be taken into account. -- Abstract: To fulfill the requirements for ITER in-vessel magnetic diagnostics, several coil prototypes have been developed, aiming at minimizing the disturbing effects of temperature gradients and radiation induced phenomena. As a first step in the radiation resistance testing of these prototypes, an in-situ high dose rate gamma radiation test on a selection of prototypes was performed. The aim of this test was to get a first experimental feedback regarding the behavior of the pick-up coil prototypes under radiation. Five prototypes (a coil wound with glass-insulated copper wire, two LTCC coils and two HTCC coils) were irradiated at a dose rate of 46 kGy/h up to a total dose of 72 MGy and at a temperature of 50 °C. During the irradiation, the continuity resistances and the insulation resistances were continuously measured. Before and after irradiation reference data were recorded as a function of temperature (from 30 °C to 130 °C). This paper includes the results of the temperature and irradiation tests and a discussion of the behavior of the prototype coils in terms of electrical and mechanical properties

ITER...ation

International Nuclear Information System (INIS)

Troyon, F.

1997-01-01

Recurrent attacks against ITER, the new generation of tokamak are a mix of political and scientific arguments. This short article draws a historical review of the European fusion program. This program has allowed to build and manage several installations in the aim of getting experimental results necessary to lead the program forwards. ITER will bring together a fusion reactor core with technologies such as materials, superconductive coils, heating devices and instrumentation in order to validate and delimit the operating range. ITER will be a logical and decisive step towards the use of controlled fusion. (A.C.)

Radwaste management aspects of the test blanket systems in ITER

Energy Technology Data Exchange (ETDEWEB)

Laan, J.G. van der, E-mail: JaapG.vanderLaan@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Canas, D. [CEA, DEN/DADN, centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Chaudhari, V. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India); Iseli, M. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Kawamura, Y. [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Petit, P. [European Commission, DG ENER, Brussels (Belgium); Pitcher, C.S.; Torcy, D. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Ugolini, D. [Fusion for Energy, Barcelona (Spain); Zhang, H. [China Nuclear Energy Industry Corporation, Beijing 100032 (China)

2016-11-01

Highlights: • Test Blanket Systems are operated in ITER to test tritium breeding technologies. • The in-vessel parts of TBS become radio-active during the ITER nuclear phase. • For each TBM campaign the TBM, its shield and the Pipe Forests are removed. • High tritium contents and novel materials are specific TBS radwaste features. • A preliminary assessment confirmed RW routing, provided its proper conditioning. - Abstract: Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to nuclear operation. The significant fusion neutron production will give rise to nuclear heating and tritium breeding in the in-vessel part of the TBS. The management of the activated and tritiated structures of the TBS from operation in ITER is described. The TBS specific features like tritium breeding and power conversion at elevated temperatures, and the use of novel materials require a dedicated approach, which could be different to that needed for the other ITER equipment.

ITER: a technology test bed for a fusion reactor

International Nuclear Information System (INIS)

Huguet, M.; Green, B.J.

1996-01-01

The ITER Project aims to establish nuclear fusion as an energy source that has potential safety and environmental advantages, and to develop the technologies required for a fusion reactor. ITER is a collaborative project between the European Union, Japan, the Russian Federation and the United States of America. During the current phase of the Project, an R and D programme of about 850 million dollars is underway to develop the technologies required for ITER. This technological effort should culminate in the construction of the components and systems of the ITER machine and its auxiliaries. The main areas of technological development include the first wall and divertor technology, the blanket technology and tritium breeding, superconducting magnet technology, pulsed power technology and remote handling. ITER is a test bed and an essential step to establish the technology of future fusion reactors. Many of the ITER technologies are of potential interest to other fields and their development is expected to benefit the industries involved. (author)

A preliminary magnetic study of Sawa lake sediments, Southern Iraq

Science.gov (United States)

Ameen, Nawrass

2016-04-01

A preliminary magnetic study combined with chemical analyses was carried out in Sawa Lake in Al-Muthanna province, southern Iraq, about 22 km south west of Samawa city (31°18'48.80"N, 45°0'25.25"E). The lake is about 4.74 km length, 1.75 km width and 5.5 m height, it is surrounded by a salt rim which is higher than the lake water by about 2.8 m and sea water by about 18.5 m (Naqash et al., 1977 in Hassan, 2007). The lake is an elongated closed basin with no surface water available to it, it may be fed by groundwater of the Euphrates and Dammam aquifers through system of joints and cracks. This study aims to investigate the concentrations of selected heavy metals as pollutants and magnetic susceptibility (MS) and other magnetic properties of sediment samples from fifty sites collected from the bottom of the lake, the study area lies in an industrial area. The results show spatial variations of MS with mean value of about 4.58 x 10-8 m3 kg-1. Scanning electron microscopy and magnetic mineralogy parameters indicate the dominance of soft magnetic phase like magnetite and presence of hard magnetic phase like hematite. Spatial variations of MS combined with the concentrations of heavy metals suggests the efficiency of magnetic methods as effective, inexpensive and non-time consuming method to outlining the heavy metal pollution. References: Hassan W.F., 2007. The Physio-chemical characteristic of Sawa lake water in Samawa city-Iraq. Marine Mesopotamica, 22(2), 167-179.

ITER and the fusion reactor: status and challenge to technology

International Nuclear Information System (INIS)

Lackner, K.

2001-01-01

Fusion has a high potential, but requires an integrated physics and technology effort without precedence in non-military R and D, the basic physics feasibility demonstration will be concluded with ITER, although R and D for efficiency improvement will continue. The essential technological issues remaining at the start of ITER operation concern materials questions: first wall components and radiation tolerant (low activation materials). This paper comprised just the copy of the slides presentation with the following subjects: magnetic confinement fusion, the Tokamak, progress in Tokamak performance, ITER: its geneology, physics basis-critical issues, cutaway of ITER-FEAT, R and D - divertor cassette (L-5), differences power plant-ITER, challenges for ITER and fusion plants, main technological problems (plasma facing materials), structural and functional materials for fusion power plants, ferritic steels, EUROFER development, improvements beyond ferritic steels, costing among others. (nevyjel)

Design and manufacturing feasibility of ITER TBM Frame and Dummy TBMs

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung-Yoon, E-mail: byoungyoon.kim@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Marconi, Matteo [LTCalcoli, Piazza Prinetti 26/B, 23807 Merate (Italy); Maio, Pietro Alessandro Di [University of Palermo, Viale delle Scienze, Edificio 6, 90128 Palermo (Italy); Torres, Luc [CNIM, Zone Portuaire De Bregaillon, CS60208, 83507 La Seyne Sur Mer Cedex (France); Lucca, Flavio [LTCalcoli, Piazza Prinetti 26/B, 23807 Merate (Italy); Gicquel, Stefan; Merola, Mario; Giancarli, Luciano M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

2016-11-01

Highlights: • The metallic gasket sealing performance was confirmed with required compression load. • The structural assessments satisfied the RCC-MR structural requirements. • Cooling circuits were improved considering flow and draining performances. • The manufacturing feasibility was preliminary assessed. - Abstract: The operation and test of mock-ups of tritium breeding blankets relevant for a future commercial reactor is one of the goals of the ITER machine. To accomplish this goal, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are installed in three ITER equatorial ports. Each TBM and the associated shield form a TBM-set that is mechanically attached to a steel frame called TBM Frame. A Frame and two TBM-Sets form a TBM Port Plug (TBM PP). The ITER Organization is responsible for the design and manufacture of the TBM Frames and of the Dummy TBMs that could replace the TBM-sets in case they were not available. This paper describes the recent results of the design supporting analyses for the TBM Frames and Dummy TBMs that is presently in the preliminary design stage and their impact on the design.

Design and manufacturing feasibility of ITER TBM Frame and Dummy TBMs

International Nuclear Information System (INIS)

Kim, Byoung-Yoon; Marconi, Matteo; Maio, Pietro Alessandro Di; Torres, Luc; Lucca, Flavio; Gicquel, Stefan; Merola, Mario; Giancarli, Luciano M.

2016-01-01

Highlights: • The metallic gasket sealing performance was confirmed with required compression load. • The structural assessments satisfied the RCC-MR structural requirements. • Cooling circuits were improved considering flow and draining performances. • The manufacturing feasibility was preliminary assessed. - Abstract: The operation and test of mock-ups of tritium breeding blankets relevant for a future commercial reactor is one of the goals of the ITER machine. To accomplish this goal, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are installed in three ITER equatorial ports. Each TBM and the associated shield form a TBM-set that is mechanically attached to a steel frame called TBM Frame. A Frame and two TBM-Sets form a TBM Port Plug (TBM PP). The ITER Organization is responsible for the design and manufacture of the TBM Frames and of the Dummy TBMs that could replace the TBM-sets in case they were not available. This paper describes the recent results of the design supporting analyses for the TBM Frames and Dummy TBMs that is presently in the preliminary design stage and their impact on the design.

Development of pellet injection systems for ITER

International Nuclear Information System (INIS)

Combs, S.K.; Gouge, M.J.; Baylor, L.R.

1995-01-01

Oak Ridge National Laboratory (ORNL) has been developing innovative pellet injection systems for plasma fueling experiments on magnetic fusion confinement devices for about 20 years. Recently, the ORNL development has focused on meeting the complex fueling needs of the International Thermonuclear Experimental Reactor (ITER). In this paper, we describe the ongoing research and development activities that will lead to a ITER prototype pellet injector test stand. The present effort addresses three main areas: (1) an improved pellet feed and delivery system for centrifuge injectors, (2) a long-pulse (up to steady-state) hydrogen extruder system, and (3) tritium extruder technology. The final prototype system must be fully tritium compatible and will be used to demonstrate the operating parameters and the reliability required for the ITER fueling application

Status and issues of the European contribution to ITER

International Nuclear Information System (INIS)

Bindslev, H.

2015-01-01

Highlights: • We describe the technical status of F4E's contributions to the ITER International Fusion Energy Project. • The foundations of the ITER Tokamak Complex have been completed. • We describe the production of the Toroidal Field coils and the achieved accuracy. • The first stage of ITER's pre-qualification programme for the ITER first wall panels was completed. • Technical developments for several other ITER components are described. - Abstract: Fusion for Energy (F4E), on behalf of Europe, is responsible for the procurement of most of the high-technology items for the ITER device. This paper provides an overview of the technical status of Europe's contributions to ITER and the related challenges. In particular, we report on progress in the construction of the buildings at the Cadarache site, the fabrication of the superconducting magnets and the vacuum vessel and the testing and qualification of the in-vessel components (first wall and divertor). The status of the design and development of the additional heating systems and the test blanket modules will also be described.

Preliminary study of an integral harmonic analysis magnetic field measurement system for long SSC magnets

International Nuclear Information System (INIS)

Green, M.I.

1991-04-01

We described the research and development required to design and build a prototype system capable of making integrated magnetic multipole measurements of warm and cryogenic 50 mm bore SSC dipole and quadrupole magnets utilizing a warm probe in a warm finger. Our experience and some preliminary studies indicate that it is highly unlikely that a 16 meter long probe can be fabricated that will have a twist below several milliradians at any temperature. Consequently we describe a segmented 16 meter long probe for which we intend to calibrate the phase of each segment to within 0.1 milliradians. The data for all segments will be acquired simultaneously, and integrated data will be generated from the vector sums of the individual segments. The calibration techniques and instrumentation required to implement this system are described. The duration of an integral measurement at one current is less than ten seconds, which is three orders of magnitude shorter than that required by the mole technique presently being used. The system is based on an extrapolation of the techniques used at LBL to measure cryogenic 1 meter models of SSC magnets with a cryogenic probe. 3 refs., 3 figs

Status of Design and Manufacturing of ITER 1st batch Assembly Tools

Energy Technology Data Exchange (ETDEWEB)

Bae, Jin Ho; Nam, Kyoun Go; Chung, Si Kun; Ha, Min Su [ITER Korea National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Geun Hong [ITER Organization, St Paul lez Durance (France)

2016-05-15

The ITER tokamak assembly tools are purpose-built and specially designed to complete the ITER tokamak machine which includes; Vacuum Vessel (VV), VV Thermal Shield (VVTS), Toroidal Field Coil (TFC) and other components contained in the cryostat. KODA has carried out the preliminary and final design of these assembly tools. This paper shows that the current status, first quarter of the 2016, including manufacturing of ITER 1st batch assembly tools and briefly summarized the design process through design work of Sector Sub-assembly Tool (SSAT) that is most important tool representing ITER 1st batch assembly tools. KODA (Korea Domestic Agency) should provide 128 kinds of the purpose-built assembly tools for ITER Tokamak machine, and the ITER 1st batch assembly tools are split into 3 groups. The FDR for Group A was performed in December 2014, and design of SSAT has been verified by FE analysis and engineering calculation using EN cords. The SSAT is now under manufacturing phase to meet the ITER milestone. After factory acceptance test of SSAT on end of 2016, the 1st SSAT will be delivered and arrived in ITER site on second quarter of the 2017.

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

Energy Technology Data Exchange (ETDEWEB)

Rizzo, Enrico, E-mail: enrico.rizzo@kit.edu [Institute for Technical Physics, Karlsruhe Institute of Technology, 76344 Eggenstein-Leopoldshafen (Germany); Bauer, Pierre [ITER Organization, Cadarache (France); Heller, Reinhard [Institute for Technical Physics, Karlsruhe Institute of Technology, 76344 Eggenstein-Leopoldshafen (Germany); Richard, Laura Savoldi; Zanino, Roberto [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy)

2013-12-15

Highlights: • A global, predictive picture of the ITER HTS current lead is not yet available. • A predictive 1-D, steady state thermal-hydraulic analysis of the full length HTS current leads has been performed. • For the heat exchanger, correlations previously derived by the same authors have been used. • The results have been compared with the ITER relevant requirements. • According to our results, the ITER HTS current leads will fulfill the requirements. -- Abstract: The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R and D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.

Towards fully authentic modelling of ITER divertor plasmas

International Nuclear Information System (INIS)

Maddison, G.P.; Hotston, E.S.; Reiter, D.; Boerner, P.

1991-01-01

Ignited next step tokamaks such as NET or ITER are expected to use a poloidal magnetic divertor to facilitate exhaust of plasma particles and energy. We report a development coupling together detailed computational models for both plasma and recycled neutral particle transport processes, to produce highly detailed and consistent design solutions. A particular aspect is involvement of an accurate specification of edge magnetic geometries, determined by an original equilibrium discretisation code, named LINDA. Initial results for a prototypical 22MA ITER double-null configuration are presented. Uncertainties in such modelling are considered, especially with regard to intrinsic physical scale lengths. Similar results produced with a simple, analytical treatment of recycling are also compared. Finally, a further extension allowing true oblique target sections is anticipated. (author) 8 refs., 5 figs

ITER Safety and Licensing

International Nuclear Information System (INIS)

Girard, J-.P; Taylor, N.; Garin, P.; Uzan-Elbez, J.; GULDEN, W.; Rodriguez-Rodrigo, L.

2006-01-01

The site for the construction of ITER has been chosen in June 2005. The facility will be implemented in Europe, south of France close to Marseille. The generic safety scheme is now under revision to adapt the design to the host country regulation. Even though ITER will be an international organization, it will have to comply with the French requirements in the fields of public and occupational health and safety, nuclear safety, radiation protection, licensing, nuclear substances and environmental protection. The organization of the central team together with its partners organized in domestic agencies for the in-kind procurement of components is a key issue for the success of the experimentation. ITER is the first facility that will achieve sustained nuclear fusion. It is both important for the experimental one-of-a-kind device, ITER itself, and for the future of fusion power plants to well understand the key safety issues of this potential new source of energy production. The main safety concern is confinement of the tritium, activated dust in the vacuum vessel and activated corrosion products in the coolant of the plasma-facing components. This is achieved in the design through multiple confinement barriers to implement the defence in depth approach. It will be demonstrated in documents submitted to the French regulator that these barriers maintain their function in all postulated incident and accident conditions. The licensing process started by examination of the safety options. This step has been performed by Europe during the candidature phase in 2002. In parallel to the final design, and taking into account the local regulations, the Preliminary Safety Report (RPrS) will be drafted with support of the European partner and others in the framework of ITER Task Agreements. Together with the license application, the RPrS will be forwarded to the regulatory bodies, which will launch public hearings and a safety review. Both processes must succeed in order to

Comparative study of low and high aspect ratio devices for ITER design options

International Nuclear Information System (INIS)

Sugihara, Masayoshi; Tada, Eisuke; Shimomura, Yasuo; Tsunematsu, Toshihide; Nishio, Satoshi; Nakazato, Toshiko; Murakami, Yoshiki; Koizumi, Koichi

1992-09-01

Comparative study on the plasma performance and the engineering characteristics of low and high aspect ratio devices for ITER (International Thermonuclear Experimental Reactor) design option is done to examine quantitatively the expected merit and demerit of high aspect ratio device on steady state operation. Device parameters of aspect ratio A=3 and 4 are chosen based on ITER-power scaling law. Improvement of steady state operation with A=4 is found only moderate. Reduction of stability margin in vertical instability is about 20% and plasma elongation must be decreased from 2 down to about 1.8 to recover this reduction of stability margin with A=4. If such lower elongation is employed, single null divertor configuration should be employed to reduce the capacity of poloidal field system. Detailed 2D divertor code calculation shows that peak heat load per unit area of A=4 device with SN configuration increases compared with A=3 device with DN configuration, contrary to the predictions so far made. Preliminary engineering studies indicate that A=4 device would have less space for handling the in-vessel components and doubled toroidal field magnet weight and winding length, and hence is less desirable when compared with the present ITER design (A=3). Based on these examinations, it is concluded that high aspect ratio device does not have remarkable advantage than low aspect ratio device, and the latter device has similar capability for the prospect of future commercial reactor to the former device. (J.P.N.)

Physics fundamentals for ITER

International Nuclear Information System (INIS)

Rosenbluth, M.N.

1999-01-01

The design of an experimental thermonuclear reactor requires both cutting-edge technology and physics predictions precise enough to carry forward the design. The past few years of worldwide physics studies have seen great progress in understanding, innovation and integration. We will discuss this progress and the remaining issues in several key physics areas. (1) Transport and plasma confinement. A worldwide database has led to an 'empirical scaling law' for tokamaks which predicts adequate confinement for the ITER fusion mission, albeit with considerable but acceptable uncertainty. The ongoing revolution in computer capabilities has given rise to new gyrofluid and gyrokinetic simulations of microphysics which may be expected in the near future to attain predictive accuracy. Important databases on H-mode characteristics and helium retention have also been assembled. (2) Divertors, heat removal and fuelling. A novel concept for heat removal - the radiative, baffled, partially detached divertor - has been designed for ITER. Extensive two-dimensional (2D) calculations have been performed and agree qualitatively with recent experiments. Preliminary studies of the interaction of this configuration with core confinement are encouraging and the success of inside pellet launch provides an attractive alternative fuelling method. (3) Macrostability. The ITER mission can be accomplished well within ideal magnetohydrodynamic (MHD) stability limits, except for internal kink modes. Comparisons with JET, as well as a theoretical model including kinetic effects, predict such sawteeth will be benign in ITER. Alternative scenarios involving delayed current penetration or off-axis current drive may be employed if required. The recent discovery of neoclassical beta limits well below ideal MHD limits poses a threat to performance. Extrapolation to reactor scale is as yet unclear. In theory such modes are controllable by current drive profile control or feedback and experiments should

U.S. Contributions to ITER

International Nuclear Information System (INIS)

Sauthoff, Ned R.

2005-01-01

The United States participates in the ITER project and program to enable the study of the science and technology of burning plasmas, a key programmatic element missing from the world fusion program. The 2003 U.S. decision to enter the ITER negotiations followed an extensive series of community and governmental reviews of the benefits, readiness, and approaches to the study of burning plasmas. This paper describes both the technical and the organizational preparations and plans for U.S. participation in the ITER construction activity: in-kind contributions, staff contributions, and cash contributions as well as supporting physics and technology research. Near-term technical activities focus on the completion of R and D and design and mitigation of risks in the areas of the central solenoid magnet, shield/blanket, diagnostics, ion cyclotron system, electron cyclotron system, pellet fueling system, vacuum system, tritium processing system, and conventional systems. Outside the project, the U .S. is engaged in preparations for the test blanket module program. Organizational activities focus on preparations of the project management arrangements to maximize the overall success of the ITER Project; elements include refinement of U.S. directions on the international arrangements, the establishment of the U.S. Domestic Agency, progress along the path of the U.S. Department of Energy's Project Management Order, and overall preparations for commencement of the fabrication of major items of equipment and for provision of staff and cash as specified in the upcoming ITER agreement

Recommendations for a cryogenic system for ITER [International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Slack, D.S.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Soviet Union, and the United States. ITER will be a large machine requiring up to 100 kW of refrigeration at 4.5 K to cool its superconducting magnets. Unlike earlier fusion experiments, the ITER cryogenic system must handle pulse loads constituting a large percentage of the total load. These come from neutron heating during a fusion burn and from ac losses during ramping of current in the PF (poloidal field) coils. This paper presents a conceptual design for a cryogenic system that meets ITER requirements. It describes a system with the following features: Only time-proven components are used. The system obtains a high efficiency without use of cold pumps or other developmental components. High reliability is achieved by paralleling compressors and expanders and by using adequate isolation valving. The problem of load fluctuations is solved by a simple load-leveling device. The cryogenic system can be housed in a separate building located at a considerable distance from the ITER core, if desired. The paper also summarizes physical plant size, cost estimates, and means of handling vented helium during magnet quench. 4 refs., 4 figs., 3 tabs

Qualification of the US Made Conductors for ITER TF Magnet System

International Nuclear Information System (INIS)

Martovetsky, Nicolai N.; Hatfield, Daniel R.; Miller, John R.; Bruzzone, P.; Stepanov, B.; Seber, B.

2010-01-01

The US Domestic Agency (USDA) is one of the six suppliers of the Toroidal Field (TF) conductor for the International Thermonuclear Experimental Reactor (ITER). In order to qualify conductors according to ITER requirements we prepared several lengths of the CICC and short samples for testing in the SULTAN facility in CRPP, Switzerland. We also fully characterized the strands that were used in these SULTAN samples. Fabrication experience and test results are presented and discussed.

Novel aspects of plasma control in ITER

Energy Technology Data Exchange (ETDEWEB)

Humphreys, D.; Jackson, G.; Walker, M.; Welander, A. [General Atomics P.O. Box 85608, San Diego, California 92186-5608 (United States); Ambrosino, G.; Pironti, A. [CREATE/University of Naples Federico II, Napoli (Italy); Vries, P. de; Kim, S. H.; Snipes, J.; Winter, A.; Zabeo, L. [ITER Organization, St. Paul Lez durance Cedex (France); Felici, F. [Eindhoven University of Technology, Eindhoven (Netherlands); Kallenbach, A.; Raupp, G.; Treutterer, W. [Max-Planck Institut für Plasmaphysik, Garching (Germany); Kolemen, E. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Lister, J.; Sauter, O. [Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland); Moreau, D. [CEA, IRFM, 13108 St. Paul-lez Durance (France); Schuster, E. [Lehigh University, Bethlehem, Pennsylvania (United States)

2015-02-15

ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

Preliminary Planck constant measurements via UME oscillating magnet Kibble balance

Science.gov (United States)

Ahmedov, H.; Babayiğit Aşkın, N.; Korutlu, B.; Orhan, R.

2018-06-01

The UME Kibble balance project was initiated in the second half of 2014. During this period we have studied the theoretical aspects of Kibble balances, in which an oscillating magnet generates AC Faraday’s voltage in a stationary coil, and constructed a trial version to implement this idea. The remarkable feature of this approach is that it can establish the link between the Planck constant and a macroscopic mass by one single experiment in the most natural way. Weak dependences on variations of environmental and experimental conditions, small size, and other useful features offered by this novel approach reduce the complexity of the experimental set-up. This paper describes the principles of the oscillating magnet Kibble balance and gives details of the preliminary Planck constant measurements. The value of the Planck constant determined with our apparatus is \\boldsymbol{h}/{{\\boldsymbol{h}}\\boldsymbol 90}={1}{.000} {004}~ , with a relative standard uncertainty of 6 ppm.

Some topics in matrix iterative analysis

International Nuclear Information System (INIS)

Khandekar, D.C.; Menon, S.V.G.; Sahni, D.C.

1984-01-01

This report deals with the general theory of matrix iterative analysis. The contents of the report are presented in the form of lecture notes primarily because the report is an outcome of a series of lectures delivered in the Theoretical Reactor Physics Section of the Bhabha Atomic Research Centre, Bombay. The first six lectures are devoted to the mathematical preliminaries needed to fully understand the subject. The remaining lectures provide an introduction to various iteractive methods and their intercomparison. (author)

Manufacture and testing of the CTB&SBB thermal shield for the ITER magnet feeder system

Energy Technology Data Exchange (ETDEWEB)

Lu, Kun; Liu, Chen; Song, Yuntao; Feng, Hansheng; Ding, Kaizhong, E-mail: kzding@ipp.ac.cn; Wang, Tanbin; Ji, Hui

2015-10-15

The system of International Thermonuclear Experimental Reactor (ITER) feeders is responsible for the power, helium cooling, and instrumentation of the magnets of the coil terminal box and S-bend box (CTB&SBB) thermal shield outside the cryostat. An 80-K rectangular Al thermal shield is hung inside the CTB&SBB to reduce the thermal radiation heat loads of 4.5-K helium. The American Society of Interventional Pain Physicians (ASIPP) will supply all the 31 sets of feeders for ITER. A manufactured prototype of CTB&SBB thermal shield is first quality-tested before the commencement of the series production. First, a detailed configuration of the rectangular Al thermal shield is presented in this article. The paper also presents more information on the manufacturing process of the thermal shield, especially the welding process, the procedure for ensuring good weld quality, and the use of a specially designed tool to ensure <5-mm deformation on such a 7.3-m-long thermal shield during welding. In addition, the cold test and results, including the cooling process with 13-bar and 17.5-g/s 80-K He gas, and the temperature distribution on different panels of the thermal shield are presented. The whole process of manufacture and testing lays a good foundation for the series production of the thermal shield.

Risk mitigation strategy for the ITER electron cyclotron upper port launcher

International Nuclear Information System (INIS)

Goede, A.P.H.; Bongers, W.A.; Elzendoorn, B.S.Q.; Graswinckel, M.F.; Baar, M.R. de

2010-01-01

A basic requirement for ITER equipment to meet is a high level of reliability, because ITER operation time is precious and radioactive operation leaves limited scope for repair. In order to reduce the risk of failure during ITER operation an effective risk mitigation strategy is necessary. This paper presents such strategy for the ITER electron cyclotron upper port launcher (ECUPL). A preliminary ECUPL risk analysis identifies possible failure modes. A probabilistic risk assessment quantifies the risk of failure using a 4 x 4 impact-likelihood matrix. Impact is quantified through technical, cost and schedule elements. Likelihood depends on the risk mitigation strategy adopted. A cost benefit analysis determines the most cost effective risk mitigation strategy. An essential element in risk mitigation is the testing of equipment prior to installation on the ITER machine. This paper argues the need for low- and highpower millimetre wave tests carried out on the fully assembled ECUPL. It presents a conceptual design for a dedicated on-site test bed that can handle 2 of 8 microwave beams at 2 MW long pulse operation.

Manufacturing progress of EDIPO. A Nb{sub 3}Sn-dipole for the ITER conductor test facility

Energy Technology Data Exchange (ETDEWEB)

Borlein, M.; Amend, J.; Theisen, W.; Walter, W. [Babcock Noell GmbH, Wuerzburg (Germany); Baker, W.; Fernandez-Cano, E.; Portone, A.; Salpietro, E. [Fusion For Energy F4E, Barcelona (Spain)

2010-05-15

ITER (International Thermonuclear Experimental Reactor) is a joint international research and development project that aims to demonstrate the scientific and technical feasibility of fusion power. For the construction of ITER and the manufacturing of its components, high quality standards must be met. Especially the coils of the magnet system - the heart of the ITER machine - are unique in size and complexity. Therefore the magnet coil manufacturing must be followed by a lot of quality measures. One of the necessary tests is the control of the conductor - to be sure that the conductor fulfils the technical performance needed for a proper magnet operation. As the conductor will experience a magnetic field of approx. 12-13 T during operation, it has to be tested within a magnetic background field. The European Dipole-magnet, called EDIPO, will be the heart of this conductor test facility which will be constructed at the CRPP Villigen (CH). Following the presentation, given in the Annual Meeting on Nuclear Technology in 2007, this paper shows the current status of the manufacturing of this complex Nb{sub 3}Sn-Magnet. At first, the design of the EDIPO shall be described. (orig.)

An overview of the ITER project

International Nuclear Information System (INIS)

Holtkamp, N.

2007-01-01

The ITER Project Team now coming together in Cadarache is currently being shaped from the old, preserving the legacy of technical know-how built up in the ITER Joint Central Team since 1992. It is particularly strong initially in the most urgent areas, related to long lead items - magnets, the main vessel and the buildings - as well as in work related to licensing. But it also incorporates new functional needs - financial, administrative, and procurement - and ties in the needs of future users during operation. Since the bulk of the procurement for ITER will be provided in kind, efforts have been strengthened to define better the share of responsibilities with the Parties' Domestic Agencies. The procurement cost sharing is being transferred into realistic technical splitting of the work, with agreements between the Parties to demonstrate production of the necessary quality, and how to handle any shortcomings. The design has evolved since originally conceived and valued 5 years ago. Design reviews of specific procurements will therefore start in September 2006 to ensure the current manufacturing and design assumptions continue to satisfy requirements. This paper reviews the current status of development of the ITER project, covering organisational and technical issues

Iterative algorithm for the volume integral method for magnetostatics problems

International Nuclear Information System (INIS)

Pasciak, J.E.

1980-11-01

Volume integral methods for solving nonlinear magnetostatics problems are considered in this paper. The integral method is discretized by a Galerkin technique. Estimates are given which show that the linearized problems are well conditioned and hence easily solved using iterative techniques. Comparisons of iterative algorithms with the elimination method of GFUN3D shows that the iterative method gives an order of magnitude improvement in computational time as well as memory requirements for large problems. Computational experiments for a test problem as well as a double layer dipole magnet are given. Error estimates for the linearized problem are also derived

Electromagnetic analysis, structural integrity and progress on mechanical design of the ITER ferromagnetic insert

Energy Technology Data Exchange (ETDEWEB)

Morimoto, M. [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan)], E-mail: masaaki_morimoto@maia.eonet.ne.jp; Ioki, K.; Terasawa, A.; Utin, Yu.; Barabash, V.; Gribov, Y. [ITER Organization, 13108 St. Paul lez Durance (France)

2009-12-15

Ferromagnetic material is used to reduce the toroidal field ripple in JFT-2M and JT-60U . In ITER, since the ferromagnetic material is inserted in the space between the double walls of ITER Vacuum Vessel (VV), it is called 'ferromagnetic inserts'. Suitable material is selected to satisfy the design requirements of ITER. The proper location and amount of the ferromagnetic inserts are optimized with the goal of reduction of the toroidal field ripple. The ferromagnetic inserts are designed to minimize electromagnetic forces acting on them. The electromagnetic forces have been calculated with the latest disruption scenarios. Magnetization forces due to magnetic fields have also been calculated. Structural integrity has been validated by a structural analysis.

Analysis of quench-vent pressures for present design of ITER [International Thermonuclear Experimental Reactor] TF [toroidal field] coils

International Nuclear Information System (INIS)

Slack, D.S.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Union of the Soviet Union, and the United States. This paper examines the effects of a quench within the toroidal field (TF) coils based on current ITER design. It is a preliminary, rough analysis. Its intent is to assist ITER designers while more accurate computer codes are being developed and to provide a check against these more rigorous solutions. Rigorous solutions to the quench problem are very complex involving three-dimensional heat transfer, extreme changes in heat capacities and copper resistivity, and varying flow dynamics within the conductors. This analysis addresses all these factors in an approximate way. The result is much less accurate than a rigorous analysis. Results here could be in error as much as 30 to 40 percent. However, it is believed that this paper can still be very useful to the coil designer. Coil pressures and temperatures vs time into a quench are presented. Rate of helium vent, energy deposition in the coil, and depletion of magnetic stored energy are also presented. Peak pressures are high (about 43 MPa). This is due to the very long vent path length (446 m), small hydraulic diameters, and high current densities associated with ITER's cable-in-conduit design. The effects of these pressures as well as the ability of the coil to be self protecting during a quench are discussed. 3 refs., 3 figs., 1 tab

An Overview Of The ITER In-Vessel Coil Systems

International Nuclear Information System (INIS)

Heitzenroeder, P.J.; Brooks, A.W.; Chrzanowski, J.H.; Dahlgren, F.; Hawryluk, R.J.; Loesser, G.D.; Neumeyer, C.; Mansfield, C.; Smith, J.P.; Schaffer, M.; Humphreys, D.; Cordier, J.J.; Campbell, D.; Johnson, G.A.; Martin, A.; Rebut, P.H.; Tao, J.O.; Fogarty, P.J.; Nelson, B.E.; Reed, R.P.

2009-01-01

ELM mitigation is of particular importance in ITER in order to prevent rapid erosion or melting of the divertor surface, with the consequent risk of water leaks, increased plasma impurity content and disruptivity. Exploitable 'natural' small or no ELM regimes might yet be found which extrapolate to ITER but this cannot be depended upon. Resonant Magnetic Perturbation has been added to pellet pacing as a tool for ITER to mitigate ELMs. Both are required, since neither method is fully developed and much work remains to be done. In addition, in-vessel coils enable vertical stabilization and RWM control. For these reasons, in-vessel coils (IVCs) are being designed for ITER to provide control of Edge Localized Modes (ELMs) in addition to providing control of moderately unstable resistive wall modes (RWMs) and the vertical stability (VS) of the plasma.

A simulation environment for ITER PCS development

International Nuclear Information System (INIS)

Walker, M.L.; Ambrosino, G.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neu, G.; Raupp, G.; Treutterer, W.; Winter, A.

2014-01-01

Highlights: • Describes task to develop simulation tool to aid development/testing of ITER PCS. • Requirements and use cases and preliminary architecture have been delivered. • Detailed design is now being developed. • Provides overview of use cases and requirements. • Provides overview of architecture and status of development. - Abstract: A simulation environment known as the Plasma Control System Simulation Platform (PCSSP), specifically designed to support development of the ITER Plasma Control System (PCS), is currently under construction by an international team encompassing a cross-section of expertise in simulation and exception handling for plasma control. The proposed design addresses the challenging requirements of supporting the PCS design. This paper provides an overview of the PCSSP project and a discussion of some of the major features of its design. Plasma control for the ITER tokamak will be significantly more challenging than for existing fusion devices. An order of magnitude greater performance (e.g. [1,2]) is needed for some types of control, which together with limited actuator authority, implies that optimized individual controllers and nonlinear saturation logic are required. At the same time, consequences of control failure are significantly more severe, which implies a conflicting requirement for robust control. It also implies a requirement for comprehensive and robust exception handling. Coordinated control of multiple competing objectives with significant interactions, together with many shared uses of actuators to control multiple variables, implies that highly integrated control logic and shared actuator management will be required. It remains a challenge for the integrated technologies to simultaneously address these multiple and often competing requirements to be demonstrated on existing fusion devices and adapted for ITER in time to support its operational schedule. We describe ways in which the PCSSP will help address

Megawatt Power Level 120 GHz Gyrotrons for ITER Start-Up

Energy Technology Data Exchange (ETDEWEB)

Choi, E M; Marchewka, C; Mastovsky, I; Shapiro, M A; Sirigiri, J R; Temkin, R J [MIT - Plasma Science and Fusion Center, NW16-186, 167 Albany Street, Cambridge, MA 02139 (United States)

2005-01-01

We report operation of a 110 GHz gyrotron with 1.67 MW of output power measured in short pulses (3{mu}s) at an efficiency of 42% in the TE{sub 22,6} mode. We also present a preliminary design of a 1 MW, 120 GHz gyrotron for ITER start-up with an efficiency greater than 50%.

Megawatt Power Level 120 GHz Gyrotrons for ITER Start-Up

International Nuclear Information System (INIS)

Choi, E M; Marchewka, C; Mastovsky, I; Shapiro, M A; Sirigiri, J R; Temkin, R J

2005-01-01

We report operation of a 110 GHz gyrotron with 1.67 MW of output power measured in short pulses (3μs) at an efficiency of 42% in the TE 22,6 mode. We also present a preliminary design of a 1 MW, 120 GHz gyrotron for ITER start-up with an efficiency greater than 50%

Fusion energy research for ITER and beyond

International Nuclear Information System (INIS)

Romanelli, Francesco; Laxaaback, Martin

2011-01-01

The achievement in the last two decades of controlled fusion in the laboratory environment is opening the way to the realization of fusion as a source of sustainable, safe and environmentally responsible energy. The next step towards this goal is the construction of the International Thermonuclear Experimental Reactor (ITER), which aims to demonstrate net fusion energy production on the reactor scale. This paper reviews the current status of magnetic confinement fusion research in view of the ITER project and provides an overview of the main remaining challenges on the way towards the realization of commercial fusion energy production in the second half of this century. (orig.)

Kazakhstan participation in International Experimental Reactor ITER Construction project. Work status and prospects

International Nuclear Information System (INIS)

Tazhibayeva, I.L.; Tukhvatullin, Sh.T.; Shestakov, V.P.; Kuznetsov, B.A.

2002-01-01

Kazakhstan takes part in ITER project in partnership with Russian Federation since the year of 1994. At present the technical stage of the project is completed and ITER Council should take a decision on the site for international reactor. Four countries such as Canada, Japan, Spain and France have offered their territories for being used as site for launching ITER construction. ITER partners started preparing new international agreement that will cover activities on construction, operation and decommissioning of ITER. It will also include the list of research and experimental work that is conducted in support of ITER project. Kazakhstan has already made an important contribution into technical stage realization of ITER project due to scientific and technical researches conducted by National Nuclear Center, by Institute of Experimental and Theoretical Physics and by JSC 'Ulba Metallurgical plant' ('UMP'). Research activity carried out for the support of ITER project is performed in accordance with the following main trends: Tritium safety (permeability and retentin of hydrogen isotopes during in-pile irradiation in various structural materials, co-deposed layers and protective coatings); Verification of computer codes (LOCA type) loss of coolant accidents modeling in ITER reactor; Investigation of liquid metal blanket of thermonuclear reactor (tritium production in lithium containing eutectics Li17Pb83 and ceramics Li 2 TiO 3 , study of tritium permeability). At present the working group of ITER project participants started introducing proposals for cost distribution and for placing the orders on reactor construction. Further Kazakhstan participation in ITER project may be in manufacturing high-tech parts and assemblies from commercial grades of beryllium. They will be used for armouring the reactor first wall, for its thermal protection and for protection of superconductor's components for magnetic systems that are at JSC U MP'. Scientific and technical support of

ITER and research works on magnetic confinement fusion

International Nuclear Information System (INIS)

Jacquinot, J.

2013-01-01

This article presents the main features of the ITER tokamak and its implications in terms of research concerning plasma instabilities, confinement and materials. It also shows how the financial and technological responsibilities have been shared between the 7 partners (European Union, China, South-Korea, Russia, Japan, India and United-States) and also gives a progress update of the fabrication of the components at the date of September 2012

Eight meeting of the ITER diagnostic expert group

International Nuclear Information System (INIS)

Costley, A.E.; Young, K.M.

1998-01-01

The 8. Meeting of the ITER Diagnostics Expert Group which was held in San Diego, February 1998 had two main technical goals: to discuss the status and plans for developing kinetic control, and to review the current status of the design of the magnetic system

Design and analysis of breeding blanket with helium cooled solid breeder for ITER-TBM

International Nuclear Information System (INIS)

Yuan Tao; Feng Kaiming; Chen Zhi; Wang Xiaoyu

2007-01-01

Test blanket module (TBM) is one of important components in ITER. Some of related blanket technologies of future fusion, such as tritium self-sufficiency, the exaction of high-grade heat, design criteria and safety requirements and environmental impacts, will be demonstrated in ITER-TBM. In ITER device, the three equatorial ports have allocated for TBM testing. China had proposed to develop independently the ITER-TBM with helium cooled solid breeder in 12th meeting of test blanket workgroup (TBWG-12). In this work, the preliminary design and analysis for Chinese HCSB TBM will be carried out. The TBM must be contains the function of the first wall, breeding blanket, shield and structure. Finally, in the period of preliminary investigation, HCSB TBM design adopt modularization concept which is helium as coolant and tritium purge gas, ferritic/martensitic steel as structural material, Lithium orthosilicate (Li 4 SiO 4 ) as tritium breeder, beryllium pebble as neutron multiplier. TBM is allocated in standard vertical frame port. HCSB TBM consist of first wall, backplate, breeding sub-modules, caps, grid and support plate, and breeding sub-modules is arranged by layout of 2 x 6 in blanket box. In this paper, main components of HCSB TBM will be described in detail, also performance analysis of main components have been completed. (authors)

ETR/ITER systems code

Energy Technology Data Exchange (ETDEWEB)

Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

1988-04-01

A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

ETR/ITER systems code

International Nuclear Information System (INIS)

Barr, W.L.; Bathke, C.G.; Brooks, J.N.

1988-04-01

A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs

Iterative Reconstruction Methods for Hybrid Inverse Problems in Impedance Tomography

DEFF Research Database (Denmark)

Hoffmann, Kristoffer; Knudsen, Kim

2014-01-01

For a general formulation of hybrid inverse problems in impedance tomography the Picard and Newton iterative schemes are adapted and four iterative reconstruction algorithms are developed. The general problem formulation includes several existing hybrid imaging modalities such as current density...... impedance imaging, magnetic resonance electrical impedance tomography, and ultrasound modulated electrical impedance tomography, and the unified approach to the reconstruction problem encompasses several algorithms suggested in the literature. The four proposed algorithms are implemented numerically in two...

Results on the ITER Technology R and D

International Nuclear Information System (INIS)

1999-01-01

The ITER Engineering Design Activities (EDA) have passed their originally planned six years by approval of the ITER Final Design Report at a meeting of the ITER Council held in July, 1998. The four Parties (EU, Japan, Russia, and USA) had hoped to make a decision for its construction by end of the EDA. However, the financial environment of these Parties were not optimistic to directly start construction of the device scooped in the Report. The ITER Technology R and D has been conducted by cooperation of these four Parties to provide data base and demonstrate technical feasibility on the ITER design. It contains, not only component technologies on tokamak reactor core, but also peripheral system technologies such as heating and current drive technique, remote maintenance technique, tritium technology, fuel air-in-taking/-exhausting technique, measurement diagnosis element technique, safety, and so on. Above all, seven large R and D projects are identified to demonstrate technical feasibility of manufacturing and system tests. They were planned to have scales capable of extrapolating to the ITER and of carrying out by joint efforts of a plural Parties. These projects were relating to superconducting magnet technology; vacuum vessel technology, blanket technology, divertor technology, and remote maintenance technology, among which three projects were promoted under leading of Japan. This report was prepared so as to enable to understand outline of results obtained under the seven projects on the ITER Technology R and D. (G.K.)

Failure of Magnetic Fluid Seals Operating in Water: Preliminary Conclusions

Directory of Open Access Journals (Sweden)

Matuszewski Leszek

2017-06-01

Full Text Available The article analyses properties of magnetic fluid seals installed in rotary sealing nodes which operate in the utility water environment. Seals of this type have been examined as a possible solution to the problem with ship manoeuvring propulsion sealing. The present analysis bases on laboratory durability tests of magnetic fluid seals exposed to longterm utility water loads, at different water pressures and shaft revolutions. The basic seal durability criterion was the number of revolutions made by the sealing node shaft until the appearance of water tightness loss (leakage. It was found that the main factor leading to the wear of the seal is the relative speed of the magnetic fluid with respect to that of the utility water, and this process is heavily affected by the pressure acting on the seal. The reported test results are presented in the form of diagrams showing the seal durability (time until water tightness loss as a function of rotational speed. The curves shown in the diagrams are regular, with two different rotational speed ranges: the highspeed range, when the tightness loss is relatively fast, and the low-speed range, with a clear tendency to prolong the seal lifetime. These diagrams were given the name of durability curves of the MF seal operating in water. The results of the performed tests suggest formal similarity between the experimental data distribution concerning tightness loss processes occurring in magnetic fluid seals operating in water environment and metal fatigue processes. The article proposes a preliminary simplified durability model to describe the examined phenomenon.

Plasma cleaning of ITER edge Thomson scattering mock-up mirror in the EAST tokamak

Science.gov (United States)

Yan, Rong; Moser, Lucas; Wang, Baoguo; Peng, Jiao; Vorpahl, Christian; Leipold, Frank; Reichle, Roger; Ding, Rui; Chen, Junling; Mu, Lei; Steiner, Roland; Meyer, Ernst; Zhao, Mingzhong; Wu, Jinhua; Marot, Laurent

2018-02-01

First mirrors are the key element of all optical and laser diagnostics in ITER. Facing the plasma directly, the surface of the first mirrors could be sputtered by energetic particles or deposited with contaminants eroded from the first wall (tungsten and beryllium), which would result in the degradation of the reflectivity. The impurity deposits emphasize the necessity of the first mirror in situ cleaning for ITER. The mock-up first mirror system for ITER edge Thomson scattering diagnostics has been cleaned in EAST for the first time in a tokamak using radio frequency capacitively coupled plasma. The cleaning properties, namely the removal of contaminants and homogeneity of cleaning were investigated with molybdenum mirror insets (25 mm diameter) located at five positions over the mock-up plate (center to edge) on which 10 nm of aluminum oxide, used as beryllium proxy, were deposited. The cleaning efficiency was evaluated using energy dispersive x-ray spectroscopy, reflectivity measurements and x-ray photoelectron spectroscopy. Using argon or neon plasma without magnetic field in the laboratory and with a 1.7 T magnetic field in the EAST tokamak, the aluminum oxide films were homogeneously removed. The full recovery of the mirrors’ reflectivity was attained after cleaning in EAST with the magnetic field, and the cleaning efficiency was about 40 times higher than that without the magnetic field. All these results are promising for the plasma cleaning baseline scenario of ITER.

ITER [International Thermonuclear Experimental Reactor] reactor building design study

International Nuclear Information System (INIS)

Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

Guidelines for Remote Handling Maintenance of ITER Neutral Beam Components

International Nuclear Information System (INIS)

Cordier, J.-J.; Hemsworth, R.; Bayetti, P.

2006-01-01

Remote handling maintenance of ITER components is one of the main challenges of the ITER project. This type of maintenance shall be operational for the nuclear phase of exploitation of ITER, and be considered at a very early stage since it significantly impacts on the components design, interfaces management and integration business. A large part of the R/H equipment will be procured by the EU partner, in particular the whole Neutral Beam Remote Handling (RH) equipment package. A great deal of work has already been done in this field during the EDA phase of ITER project, but improvements and alternative option that are now proposed by ITER lead to added RH and maintenance engineering studies. The Neutral Beam Heating -and- Current Drive system 1 is being revisited by the ITER project. The vertical maintenance scheme that is presently considered by ITER, may significantly impact on the reference design of the Neutral Beam (NB) system and associated components and lead to new design of the NB box itself. In addition, revision of both NB cell radiation level zoning and remote handling classification of the beam line injector will also significantly impact on components design and maintenance. Based on the experience gained on the vertical maintenance scheme, developed in detail for the ITER Neutral Beam Test Facility 2 to be built in Europe in a near future, guidelines for the revision of the design and preliminary feasibility study of the remote handling vertical maintenance scheme of beam line components are described in the paper. A maintenance option for the SINGAP3 accelerator is also presented. (author)

Lower hybrid current drive at ITER-relevant high plasma densities

International Nuclear Information System (INIS)

Cesario, R.; Amicucci, L.; Cardinali, A.; Castaldo, C.; Marinucci, M.; Panaccione, L.; Pericoli-Ridolfini, V.; Tuccillo, A. A.; Tudisco, O.; Calabro, G.

2009-01-01

Recent experiments indicated that a further non-inductive current, besides bootstrap, should be necessary for developing advanced scenario for ITER. The lower hybrid current drive (LHCD) should provide such tool, but its effectiveness was still not proved in operations with ITER-relevant density of the plasma column periphery. Progress of the LH deposition modelling is presented, performed considering the wave physics of the edge, and different ITER-relevant edge parameters. Operations with relatively high edge electron temperatures are expected to reduce the LH || spectral broadening and, consequently, enabling the LH power to propagate also in high density plasmas ( || is the wavenumber component aligned to the confinement magnetic field). New results of FTU experiments are presented, performed by following the aforementioned modeling: they indicate that, for the first time, the LHCD conditions are established by operating at ITER-relevant high edge densities.

Technical basis for the ITER-FEAT outline design. Progress in resolving open design issues from the outline design report

International Nuclear Information System (INIS)

2000-01-01

In this publication the technical basis for the ITER-FEAT outline design is presented. It comprises the Plant Design Specifications, the Safety Principles and Environmental Criteria, the Site Requirements and Site Design Assumptions. The outline of the key features of the ITER-FEAT design includes main physical parameters and assessment, design overview and preliminary safety assessment, cost and schedule

Europe wrestles with ITER site bid

CERN Multimedia

Feder, T

2003-01-01

"The European Union is in a quandary over whether to put forward the French or Spanish site to host ITER, a $5 billion magnetic fusion experiment intended to prove the feasability of fusion energy. The decision is set for 27 November, with the final site selection, between the victorious European bid and bids from Canada and Japan, to follow within a couple of months" (1 page)

The ITER Neutral Beam Test Facility towards SPIDER operation

Science.gov (United States)

Toigo, V.; Dal Bello, S.; Gaio, E.; Luchetta, A.; Pasqualotto, R.; Zaccaria, P.; Bigi, M.; Chitarin, G.; Marcuzzi, D.; Pomaro, N.; Serianni, G.; Agostinetti, P.; Agostini, M.; Antoni, V.; Aprile, D.; Baltador, C.; Barbisan, M.; Battistella, M.; Boldrin, M.; Brombin, M.; Dalla Palma, M.; De Lorenzi, A.; Delogu, R.; De Muri, M.; Fellin, F.; Ferro, A.; Gambetta, G.; Grando, L.; Jain, P.; Maistrello, A.; Manduchi, G.; Marconato, N.; Pavei, M.; Peruzzo, S.; Pilan, N.; Pimazzoni, A.; Piovan, R.; Recchia, M.; Rizzolo, A.; Sartori, E.; Siragusa, M.; Spada, E.; Spagnolo, S.; Spolaore, M.; Taliercio, C.; Valente, M.; Veltri, P.; Zamengo, A.; Zaniol, B.; Zanotto, L.; Zaupa, M.; Boilson, D.; Graceffa, J.; Svensson, L.; Schunke, B.; Decamps, H.; Urbani, M.; Kushwah, M.; Chareyre, J.; Singh, M.; Bonicelli, T.; Agarici, G.; Garbuglia, A.; Masiello, A.; Paolucci, F.; Simon, M.; Bailly-Maitre, L.; Bragulat, E.; Gomez, G.; Gutierrez, D.; Mico, G.; Moreno, J.-F.; Pilard, V.; Chakraborty, A.; Baruah, U.; Rotti, C.; Patel, H.; Nagaraju, M. V.; Singh, N. P.; Patel, A.; Dhola, H.; Raval, B.; Fantz, U.; Fröschle, M.; Heinemann, B.; Kraus, W.; Nocentini, R.; Riedl, R.; Schiesko, L.; Wimmer, C.; Wünderlich, D.; Cavenago, M.; Croci, G.; Gorini, G.; Rebai, M.; Muraro, A.; Tardocchi, M.; Hemsworth, R.

2017-08-01

SPIDER is one of two projects of the ITER Neutral Beam Test Facility under construction in Padova, Italy, at the Consorzio RFX premises. It will have a 100 keV beam source with a full-size prototype of the radiofrequency ion source for the ITER neutral beam injector (NBI) and also, similar to the ITER diagnostic neutral beam, it is designed to operate with a pulse length of up to 3600 s, featuring an ITER-like magnetic filter field configuration (for high extraction of negative ions) and caesium oven (for high production of negative ions) layout as well as a wide set of diagnostics. These features will allow a reproduction of the ion source operation in ITER, which cannot be done in any other existing test facility. SPIDER realization is well advanced and the first operation is expected at the beginning of 2018, with the mission of achieving the ITER heating and diagnostic NBI ion source requirements and of improving its performance in terms of reliability and availability. This paper mainly focuses on the preparation of the first SPIDER operations—integration and testing of SPIDER components, completion and implementation of diagnostics and control and formulation of operation and research plan, based on a staged strategy.

Modeling of divertor particle and heat loads during application of resonant magnetic perturbation fields for ELM control in ITER

Energy Technology Data Exchange (ETDEWEB)

Schmitz, O., E-mail: o.schmitz@fz-juelich.de [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Becoulet, M. [CEA/IRFM, Cadarache, 13108 St. Paul-lez-Durance Cedex (France); Cahyna, P. [IPP AS CR, Za Slovankou 3, 18200 Prague 8 (Czech Republic); Evans, T.E. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Feng, Y. [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Frerichs, H.; Kirschner, A. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Kukushkin, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Laengner, R. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Lunt, T. [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Loarte, A.; Pitts, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Reiser, D.; Reiter, D. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Saibene, G. [Fusion for Energy Joint Undertaking, Barcelona (Spain); Samm, U. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany)

2013-07-15

First results from three-dimensional modeling of the divertor heat and particle flux pattern during application of resonant magnetic perturbation fields as ELM control scheme in ITER with the EMC3-Eirene fluid plasma and kinetic neutral transport code are discussed. The formation of a helical magnetic footprint breaks the toroidal symmetry of the heat and particle fluxes. Expansion of the flux pattern as far as 60 cm away from the unperturbed strike line is seen with vacuum RMP fields, resulting in a preferable heat flux spreading. Inclusion of plasma response reduces the radial extension of the heat and particle fluxes and results in a heat flux peaking closer to the unperturbed level. A strong reduction of the particle confinement is found. 3D flow channels are identified as a consistent reason due to direct parallel outflow from inside of the separatrix. Their radial inward expansion and hence the level of particle pump out is shown to be dependent on the perturbation level.

Prototyping and testing of the Continuous External Rogowski ITER magnetic sensor

Energy Technology Data Exchange (ETDEWEB)

Moreau, Ph., E-mail: philippe.jacques.moreau@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Le-Luyer, A.; Malard, P.; Pastor, P.; Saint-Laurent, F.; Spuig, P. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Lister, J.; Toussaint, M.; Marmillod, P.; Testa, D. [Centre de Recherches en Physique des Plasmas, EPFL (Switzerland); Peruzzo, S. [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Knaster, J. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); IFMIF EVEDA, Rokkasho (Japan); Vayakis, G.; Hughes, S.; Patel, K.M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2013-10-15

Highlights: ► ITER Continuous External Rogowski (CER) are designed for plasma and vacuum vessel current measurement. ► CER are located in the casing of Toroidal Field Coils and thus will operate at 4 K. ► The design of the sensors has been completed. ► CER prototypes have been manufactured by 2 suppliers. ► The R and D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER. -- Abstract: The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R and D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R and D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed.

Prototyping and testing of the Continuous External Rogowski ITER magnetic sensor

International Nuclear Information System (INIS)

Moreau, Ph.; Le-Luyer, A.; Malard, P.; Pastor, P.; Saint-Laurent, F.; Spuig, P.; Lister, J.; Toussaint, M.; Marmillod, P.; Testa, D.; Peruzzo, S.; Knaster, J.; Vayakis, G.; Hughes, S.; Patel, K.M.

2013-01-01

Highlights: ► ITER Continuous External Rogowski (CER) are designed for plasma and vacuum vessel current measurement. ► CER are located in the casing of Toroidal Field Coils and thus will operate at 4 K. ► The design of the sensors has been completed. ► CER prototypes have been manufactured by 2 suppliers. ► The R and D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER. -- Abstract: The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R and D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R and D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed

New proposal on the development of machine protection functions for ITER diagnostics control

International Nuclear Information System (INIS)

Yamamoto, Tsuyoshi; Yatsuka, Eiichi; Hatae, Takaki; Takeuchi, Masaki; Kitazawa, Sin-iti; Ogawa, Hiroaki; Kawano, Yasunori; Itami, Kiyoshi; Ota, Kazuya; Hashimoto, Yasunori; Nakamura, Kitaru; Sugie, Tatsuo

2016-01-01

There is a need to develop ITER instrumentation and control (I and C) systems with high reliabilities. Interlock systems that activate machine protection functions are implemented on robust wired-logic systems such as programmable logic controllers (PLCs). We herein propose a software tool that generates program code templates for the control systems using PLC logic. This tool decreases careless mistakes by developers and increases reliability of the program codes. A large-scale engineering database has been implemented in the ITER project. To derive useful information from this database, we propose adding semantic data to it using the Resource Description Framework format. In our novel proposal for the ITER diagnostic control system, a guide words generator that analyzes the engineering data by inference is applied to the hazard and operability study. We validated the methods proposed in this paper by applying them to the preliminary design for the I and C system of the ITER edge Thomson scattering system. (author)

Effect of Internal and Edge Transport Barriers in ITER Simulations

International Nuclear Information System (INIS)

Pianroj, Y.; Onjun, T.; Suwanna, S.; Picha, R.; Poolyarat, N.

2009-07-01

Full text: Predictive simulations of ITER with the presence of both an edge transport barrier (ETB) and an internal transport barrier (ITB) are carried out using the BALDUR integrated predictive modeling code. In these simulations, the boundary is taken at the top of the pedestal, where the pedestal values are described using the theory-based pedestal models. These pedestal temperature models are based on three different pedestal width scalings: magnetic and flow shear stabilization (δ α ρ ζ 2 ), flow shear stabilization (δ α Root ρ Rq), and normalized poloidal pressure (δ α R Root βθ, ped). The pedestal width scalings are combined with a pedestal pressure gradient scaling based on ballooning mode limit to predict the pedestal temperature. A version of the semi-empirical Mixed Bohm/gyro Bohm (Mixed B/gB) core transport model that includes ITB effects is used to compute the evolution of plasma profiles and plasma performance, which defined by Fusion Q factor. The results from the cases excluding and including ITB are compared. The preliminary results show the Q value resulted from ITB-excluded simulation is less than the one with ITB included

External heating and current drive source requirements towards steady-state operation in ITER

Science.gov (United States)

Poli, F. M.; Kessel, C. E.; Bonoli, P. T.; Batchelor, D. B.; Harvey, R. W.; Snyder, P. B.

2014-07-01

Steady state scenarios envisaged for ITER aim at optimizing the bootstrap current, while maintaining sufficient confinement and stability to provide the necessary fusion yield. Non-inductive scenarios will need to operate with internal transport barriers (ITBs) in order to reach adequate fusion gain at typical currents of 9 MA. However, the large pressure gradients associated with ITBs in regions of weak or negative magnetic shear can be conducive to ideal MHD instabilities, reducing the no-wall limit. The E × B flow shear from toroidal plasma rotation is expected to be low in ITER, with a major role in the ITB dynamics being played by magnetic geometry. Combinations of heating and current drive (H/CD) sources that sustain reversed magnetic shear profiles throughout the discharge are the focus of this work. Time-dependent transport simulations indicate that a combination of electron cyclotron (EC) and lower hybrid (LH) waves is a promising route towards steady state operation in ITER. The LH forms and sustains expanded barriers and the EC deposition at mid-radius freezes the bootstrap current profile stabilizing the barrier and leading to confinement levels 50% higher than typical H-mode energy confinement times. Using LH spectra with spectrum centred on parallel refractive index of 1.75-1.85, the performance of these plasma scenarios is close to the ITER target of 9 MA non-inductive current, global confinement gain H98 = 1.6 and fusion gain Q = 5.

ITER EDA newsletter. V. 2, no. 2

International Nuclear Information System (INIS)

1993-02-01

This issue of the ITER EDA (Engineering Design Activities) Newsletter contains a progress report by the Director, a report on the completion of the first activities of the Special Work Group 2 (SWG-2), a report on a magnet technical meeting, held at Naka, Japan, 26-29 January 1993, and on the US Home Team National Meeting, 25-26 January 1993

ITER PF6 double pancakes winding line

Energy Technology Data Exchange (ETDEWEB)

Du, Shuangsong [Institute of Plasma Physics, Chinese Academy of Science, Hefei (China); University of Science and Technology of China, Hefei (China); Wen, Wei, E-mail: wenwei@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Science, Hefei (China); Chen, Jin; Wu, Weiyue; Song, Yuntao; Shen, Guang [Institute of Plasma Physics, Chinese Academy of Science, Hefei (China)

2017-03-15

Highlights: • ITER PF6 double pancakes “two-in-hand” winding line layout and main parameters were introduced, main winding sequences were also included. • Main features of each winding unit include de-spooling unit, straightening unit, sandblasting and cleaning unit, bending unit, turn insulation wrapping head, rotary table and automatic control system were depicted. • PF6 double pancake winding line was commissioned with PF5 empty jacket conductor after the installation and testing of each unit, ±0.5 mm turn positioning and ±2 turn to turn deviations were achieved. - Abstract: The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. The PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the Poloidal Field coils cooperation agreement signed between ASIPP and Fusion for Energy (F4E).The ITER PF6 winding pack is composed by stacking of 9 double pancakes. Each double pancake is wound with a “two-in-hand” configuration. This paper describes the ITER PF6 double pancakes winding line, including layout and main parameters of the winding line, features of main units and the commissioning trial with PF5 empty jacket conductor.

Thermal effects of divertor sweeping in ITER

International Nuclear Information System (INIS)

Wesley, J.C.

1992-01-01

In this paper, thermal effects of magnetically sweeping the separatrix strike point on the outer divertor target of the International Thermonuclear Fusion Reactor (ITER) are calculated. For the 0. 2 Hz x ± 12 cm sweep scenario proposed for ITER operations, the thermal capability of a generic target design is found to be slightly inadequate (by ∼ 5%) to accommodate the full degree of plasma scrape-off peaking postulated as a design basis. The principal problem identified is that the 5 s sweep period is long relative to the 1. 4 s thermal time constant of the divertor target. An increase of the sweep frequency to ∼ 1 Hz is suggested: this increase would provide a power handling margin of ∼ 25% relative to present operational criteria

Second generation superconducting super collider dipole magnet cryostat design

International Nuclear Information System (INIS)

Niemann, R.C.; Bossert, R.C.; Carson, J.A.; Engler, N.H.; Gonczy, J.D.; Larson, E.T.; Nicol, T.H.; Ohmori, T.

1988-12-01

The SSC Magnet Development Program is developing accelerator dipole magnets in successive iterations. The initial iteration is complete with six full length model magnets and a thermal model having been built and tested. This initial experience along with the evolving SSC Magnet System Requirements have resulted in the second generation magnet cryostat design. It is this configuration that will be employed for the near term ongoing magnetic, thermal, string and accelerated life testing and will be the design considered for Phase I; i.e., Technology Orientation, of the SSC Magnet Industrialization Program. 5 refs., 7 figs., 1 tab

Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

International Nuclear Information System (INIS)

Sannazzaro, G.; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H.

2001-01-01

In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses

Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Sannazzaro, G. E-mail: sannazg@itereu.de; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H

2001-11-01

In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses.

A new approach of equilibrium reconstruction for ITER

International Nuclear Information System (INIS)

Imazawa, R.; Kawano, Y.; Kusama, Y.

2011-01-01

We have proposed a new approach for equilibrium reconstruction that can be applied to ITER-like burning plasmas. In this study, we have focused on carrying out equilibrium reconstruction using polarimetry, which is feasible for ITER-like burning plasmas. Polarimetry in burning plasmas is different from that in the existing tokamaks in two regards: (1) increased importance of the relativistic effects and (2) significant coupling with the Faraday and Cotton–Mouton effects. We found that when polarimetric data (orientation angle, θ, and ellipticity angle, ε, of a polarization state) are used as the constraints in the equilibrium reconstruction, the optimum weighting factors for θ and ε depend on the magnetic surfaces through which the viewing chord of polarimetry passes. We applied our approach to the operation scenarios II (S2) and IV (S4) in ITER. In the case where the viewing chords are via both the equatorial and upper ports, the measurement requirements for the accuracy of the q-profile in ITER (±10%) were satisfied in S2 and S4 when the measuring errors of θ and ε were less than 0.5° and 3°, respectively.

Project status of manufacturing of European toroidal coils ITER. Validation tests

International Nuclear Information System (INIS)

Pando, F.; Felipe, A.; Madorran, A.; Pallisa, J.; Dormicch, O.; Valle, N.; D'Urzo, C.; Marin, M.; Pesenti, P.; Lucas, J.; Moreno, N.; Bonito-Oliva, A.; Harrison, R.; Bellesia, B.; Cornelis, M.; Cornella, J.

2015-01-01

The toroidal field coils are the ITER magnets responsible for confining the plasma inside the vacuum vessel. The consortium formed by IBERDROLA Ingenieria y Construccion, ASG Superconductors y ELYTT Energy is the responsible for the supply of 10 coils that the european agency F4E has to supply for the ITER project. At present, the coils are been manufactured in La Spezia (Italy), after the qualification of all the manufacturing process and the sucessfull manufacturing of a full scale prototype. (Author)

ECRH-assisted start-up in ITER

International Nuclear Information System (INIS)

Lloyd, B.; Carolan, P.G.; Warrick, C.D.

1996-07-01

In ITER, the electric field applied for ionisation and to ramp up the plasma current may be limited to ∼ 0.3 V/m. In this case, based on established theories of the avalanche process, it is shown that ohmic breakdown in ITER is only possible over a narrow range of pressure and magnetic error field. Therefore, ECRH may be necessary to provide robust and reliable start-up. ECRH can ensure prompt breakdown over a wide range of prefill pressure and error field and can also give control over the initial time and location of breakdown. For ECRH-assisted start-up in ITER, the power and pulse length requirements are essentially determined by the need to ensure burnthrough, i.e. complete ionisation of hydrogen and the transition to high ionisation states of impurities. A 0-D code (with inclusion of some 1-D effects) has been developed to analyse burnthrough in ITER. The 0-D simulations indicate that control of the deuterium density is the key factor for ensuring successful start-up in ITER, where the effects of neutral screening and dynamic fuelling by the ex-plasma volume are also crucial. It is concluded that without ECRH, successful start-up will only be possible over a very restricted range of parameters but 3MW of absorbed ECRH power will ensure reasonably robust start-up for a broad range of conditions with beryllium impurity. In the case of carbon impurity, even with an absorbed ECRH power of 5MW one may be restricted to low prefill pressure and/or low carbon concentration for successful start-up. (Author)

Three-dimensional tolerance investigation on main ITER components

International Nuclear Information System (INIS)

Reich, J.; Chiocchio, S.; Cordier, J.-J.; Gallix, R.; Guerin, O.; Halcrow, T.

2009-01-01

ITER has to focus on all processes that ensure the permanent consistency between the requirements of ITER and the performance attributes of its components. This includes integration tolerance studies. One of the main goals of this work is to establish a sufficient tolerance scheme for all main components. The investigation in sufficient tolerance studies at a very early stage of the project will result in cost savings during the installation process. Due to the complexity of the ITER components and their several interfaces to their surroundings, it is advantageous to perform tolerance studies with a specialised tool like 'three-dimensional control systems' (3DCS) that is compatible with the ITER CATIA-V5 CAD engineering system and Enovia PRC environment. On single components (e.g. Magnet TF Coils) detailed two-dimensional tolerance schemes have been developed from the beginning. Using them as a starting point, functional or key interface tolerances have to be defined. Furthermore the tolerance studies have to consider the different configurations of each component (e.g. manufacturing stages, assembly plan, integration with surrounding, operation conditions). Especially for assembly it is necessary to analyse the final ranges which have to be achieved during the installation process. From the integration point of view, the key tolerances of all main in-cryostat ITER components have to be brought together in a complete and consistent manner.

Activation analyses updating the ITER radioactive waste assessment

International Nuclear Information System (INIS)

Pampin, R.; Zheng, S.; Lilley, S.; Na, B.C.; Loughlin, M.J.; Taylor, N.P.

2012-01-01

Highlights: ► Comprehensive updated of ITER radwaste assessment. ► Latest coupled neutronics and activation methods. ► Type A waste at shutdown decays to TFA within 100 years. ► Most type B waste at shutdown is still type B after 100 years. - Abstract: A study is reported which computes the radiation transport and activation response throughout the ITER machine and updates the ITER radioactive waste assessment using modern 3D models and up-to-date methods. The latest information on component design, maintenance, replacement schedules and materials is adopted. The radwaste classification is revised for all the major components of ITER, as well as several representative port plugs. Results include categorisation snapshots at different decay times, time histories of radiological quantities throughout the machine, and guidelines on interim decay times for components. All plasma-facing materials except tungsten are found to classify as type B due to the transmutation of their main constituents. Major contributors to the IRAS index of all materials are reported. Elemental concentration limits for type A classification of first wall and divertor materials are obtained; for the steels, only a reduction in service lifetime can reduce the waste class. Comparison of total waste amounts with earlier assessments is limited by the fact that analyses of some components are still preliminary; the trend, however, indicates a potential reduction in the total amount of waste if component segregation is demonstrated.

Project status of manufacturing of European toroidal coils ITER. Validation tests; Estado del proyecto de fabricacion de las bobinas toroidales european para el ITER. Ensayos de validacion

Energy Technology Data Exchange (ETDEWEB)

Pando, F.; Felipe, A.; Madorran, A.; Pallisa, J.; Dormicch, O.; Valle, N.; D' Urzo, C.; Marin, M.; Pesenti, P.; Lucas, J.; Moreno, N.; Bonito-Oliva, A.; Harrison, R.; Bellesia, B.; Cornelis, M.; Cornella, J.

2015-07-01

The toroidal field coils are the ITER magnets responsible for confining the plasma inside the vacuum vessel. The consortium formed by IBERDROLA Ingenieria y Construccion, ASG Superconductors y ELYTT Energy is the responsible for the supply of 10 coils that the european agency F4E has to supply for the ITER project. At present, the coils are been manufactured in La Spezia (Italy), after the qualification of all the manufacturing process and the sucessfull manufacturing of a full scale prototype. (Author)

Iterating skeletons

DEFF Research Database (Denmark)

Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost

2012-01-01

a particular skeleton ad-hoc for repeated execution turns out to be considerably complicated, and raises general questions about introducing state into a stateless parallel computation. In addition, one would strongly prefer an approach which leaves the original skeleton intact, and only uses it as a building...... block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state......Skeleton-based programming is an area of increasing relevance with upcoming highly parallel hardware, since it substantially facilitates parallel programming and separates concerns. When parallel algorithms expressed by skeletons involve iterations – applying the same algorithm repeatedly...

Plan of ITER remote experimentation center

Energy Technology Data Exchange (ETDEWEB)

Ozeki, T., E-mail: ozeki.takahisa@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Obuchi Rokkasho, Kitakami-gun, Aomori 039-3212 (Japan); Clement, S.L. [Fusion for Energy, Torres Diagonal Litoral, B3, 13/03, 08019 Barcelona (Spain); Nakajima, N. [National Institute for Fusion Science and Project Leader of IFERC, 2-166 Obuchi, Rokkasho, Kamikita-gun, Aomori 039-3212 (Japan)

2014-05-15

Plan of ITER remote experimentation center (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is described. Objectives of REC activity are (1) to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, (2) to develop the remote experiment system and verify the functions required for the remote experiment by using the Satellite Tokamak (JT-60SA) facilities in order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, and (3) to test the functions of REC and demonstrate the total system by using JT-60SA and existing other facilities in EU. Preliminary identified items to be developed are (1) Functions of the remote experiment system, such as setting of experiment parameters, shot scheduling, real time data streaming, communication by video-conference between the remote-site and on-site, (2) Effective data transfer system that is capable of fast transfer of the huge amount of data between on-site and off-site and the network connecting the REC system, (3) Storage system that can store/access the huge amount of data, including database management, (4) Data analysis software for the data viewing of the diagnostic data on the storage system, (5) Numerical simulation for preparation and estimation of the shot performance and the analysis of the plasma shot. Detailed specifications of the above items will be discussed and the system will be made in these four years in collaboration with tokamak facilities of JT-60SA and EU tokamak, experts of informatics, activities of plasma simulation and ITER. Finally, the function of REC will be tested and the total system will be demonstrated by the middle of 2017.

Sensitivity of ITER MHD global stability to edge pressure gradients

International Nuclear Information System (INIS)

Hogan, J.T.; Martynov, A.

1994-01-01

In view of the preliminary nature of boundary models for reactor tokamaks, the sensitivity to edge gradients of the global mode MHD stability of the ITER EDA configuration has been examined. The POLAR-2D equilibrium and TORUS stability codes developed by the Keldysh Institute have been used. Transport-related profiles from the PRETOR transport code (developed by the ITER Joint Central Team) and axisymmetric equilibria for these profiles from the TEQ code (L.D. Pearlstein, LLNL) were taken as a starting point for the study. These baseline profiles are found to have quite high global stability limits, in the range g(Troyon) = 4-5. The major focus of this study is to examine global mode stability assuming small variations about the baseline profiles, changing the pressure gradients near the boundary. Such changes can be expected with an improved boundary model. Reduced stability limits are found in such cases, and unstable cases with g = 2-3 are found. Thus, the assumption of ITER stability limits higher than g = 2 must be treated with caution

ITER SAFETY TASK NID-5D: Operational tritium loss and accident investigation for heat transport and water detritiation systems

International Nuclear Information System (INIS)

Kalyanam, K.M.; Fong, C.; Moledina, M.; Natalizio, A.

1995-02-01

The task objectives are to: a) determine major pathways for tritium loss during normal operation of the cooling systems and water detritiation system, b) estimate operational losses and environmental tritium releases from the heat transport and water detritiation systems of ITER, and c) prepare a preliminary Failure Modes and Effects Analysis (FMEA) for the ITER Water Detritiation System. The analysis will be used to estimate chronic environmental tritium releases (airborne and waterborne) for the ITER Cooling Systems and Water Detritiation System. The assessment will form the basis for demonstrating the acceptability of ITER for siting in the Early Safety and Environmental Characterization Study (ESECS), to be issued in early 1995. (author). 7 refs., 10 tabs., 11 figs

VDE characteristics during disruption process and its underlying acceleration mechanism in the ITER-EDA tokamak

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Nishio, Satoshi; Yoshino, Ryuji; Kessel, C.E.; Jardin, S.C.

1996-01-01

The dynamic behavior of vertical displacement events (VDEs) during a disruption and acceleration mechanisms that govern VDEs in the ITER-EDA tokamak are investigated using the Tokamak Simulation Code. A sudden plasma pressure drop (β p collapse) does not accelerate VDEs for the ITER tokamak. The geometry of the ITER resistive shell is shown to be suitable for preventing a β p collapse-induced VDE, because the magnetic field decay n-index after the β p collapse does not considerably degrade. On the other hand, it is shown that the plasma current quench (I p quench) following the energy quench can accelerate VDEs due to the vertical imbalance of the attractive force arising from the up-down asymmetric shell. The vertical location of the neutral point where the I p quench-induced VDE almost disappears is found to lie at ∼22 cm below the plasma magnetic axis of the nominal equilibrium (Z = 1.44 m). An upward and moderate I p quench-induced VDE can be expected for the nominal configuration in the ITER-EDA tokamak. It is shown that the ITER tokamak has an advantage of avoiding the fatal damage of the complicated structures of the bottom-divertor. (author)

Design considerations for ITER toroidal field coils

International Nuclear Information System (INIS)

Kalsi, S.S.; Lousteau, D.C.; Miller, J.R.

1987-01-01

The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Europe, Japan, the Union of Soviet Socialist Republics (U.S.S.R.), and the United States. This paper describes a magnetic and mechanical design methodology for toroidal field (TF) coils that employs Nb 3 Sn superconductor technology. Coil winding is sized by using conductor concepts developed for the U.S. TIBER concept. Manifold concepts are presented for the complete cooling system. Also included are concepts for the coil structural arrangement. The effects of in-plane and out-of-plane loads are included in the design considerations for the windings and case. Concepts are presented for reacting these loads with a minimum amount of additional structural material. Concepts discussed in this paper could be considered for the ITER TF coils

Study on poloidal field coil optimization and equilibrium control of ITER

International Nuclear Information System (INIS)

Shinya, Kichiro; Sugihara, Masayoshi; Nishio, Satoshi

1989-03-01

The purpose of this report is to present general features of the poloidal field coil optimization for the ITER plasma, flexibility analysis for various plasma options and some other aspect of the equilibrium control which is required for understanding plasma operation in more detail. Double null divertor plasma was selected as a main object of the optimization. Single null divertor plasma was assumed to be an alternative, because single null divertor plasma can be operational within the amounts of the total stored energy and ampere-turns of the double null divertor plasma, if it is shaped appropriately. Plasma parameters used in the present analysis are mainly those employed in the preliminary study by the Basic Device Engineering group of the ITER design team. The most part of the optimization study, however, utilizes the parameters proposed for discussion by the Japan team before starting joint design work at Garching. Plasma shape, and solenoid coil shape and size, which maximize available flux swing with reasonable amounts of the stored energy and ampere-turns, are discussed. Location and minimum number of the poloidal field coils with adequate shaping controllability were also discussed for various plasma options. Some other aspect of the equilibrium control, such as separatrix swing, moving null point operation during plasma heating and possible range of li, were evaluated and the guideline for the engineering design was proposed. Finally, fusion power output was estimated for the different pressure profiles and combinations of the average density and temperature, and the magnetic quantities of the scrape-off region was calculated to be available for the future divertor analysis. (author)

Manufacture and test of seismic bellows for ITER magnet feeder

International Nuclear Information System (INIS)

Liu, Chen; Lu, Kun; Sheng, Liang; Song, Yuntao; Su, Jinjin; Su, Man; Gung, Chenyu

2016-01-01

Highlights: • The design of the double bellows was iterated with the results of analysis based on the Expansion Joint Manufacturers Association (EJMA) standard. • The seismic bellows was tested with cyclic pressurization of the interlayer space to 2 bars absolute pressure for 5 cycles. • 200 cycles of tensile fatigue test with 90 mm of stretching from the nominal design length. • A full tensile test with 315 mm of stretching from its nominal length was conducted. • The prototype bellows was qualified for its leak tightness (less than 1 × 10 −9 Pam 3 /s of helium) at all time during the qualification test. - Abstract: This paper presents the key manufacturing and testing processes of the prototype ITER feeder seismic bellows. The design of the double bellows was iterated with the results of analysis based on the Expansion Joint Manufacturers Association (EJMA) standard. Each inner and outer bellows was supported in dedicated molds and formed by a hydraulic pressure machine rated at 800 tons. The double bellows were constructed by welding individual collars to the end flanges. The seismic bellows was tested with cyclic pressurization of the interlayer space to 2 bars absolute pressure for 5 cycles. This was followed by 200 cycles of tensile fatigue test with 90 mm of stretching from the nominal design length. After the mechanical fatigue test, a full tensile test with 315 mm of stretching from its nominal length was conducted. Helium leak tests, with the sensitivity of the helium leak detector set to 1 × 10 −9 Pa m 3 /s of helium, were performed at different stages of pressure and mechanical tests. The prototype bellows was qualified for its leak tightness at all time during the qualification test.

Manufacture and test of seismic bellows for ITER magnet feeder

Energy Technology Data Exchange (ETDEWEB)

Liu, Chen, E-mail: liuchen@ipp.ac.cn [Institute of Plasma Physics Chinese Academy of Sciences, ShuShanhu Road No. 350, Hefei (China); Lu, Kun [Institute of Plasma Physics Chinese Academy of Sciences, ShuShanhu Road No. 350, Hefei (China); Sheng, Liang [AEROSUN-TOLA Expansion Joint Co., Ltd., Jiangjun Road No. 199, Nanjing (China); Song, Yuntao; Su, Jinjin [Institute of Plasma Physics Chinese Academy of Sciences, ShuShanhu Road No. 350, Hefei (China); Su, Man; Gung, Chenyu [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul-lez-Durance (France)

2016-11-01

Highlights: • The design of the double bellows was iterated with the results of analysis based on the Expansion Joint Manufacturers Association (EJMA) standard. • The seismic bellows was tested with cyclic pressurization of the interlayer space to 2 bars absolute pressure for 5 cycles. • 200 cycles of tensile fatigue test with 90 mm of stretching from the nominal design length. • A full tensile test with 315 mm of stretching from its nominal length was conducted. • The prototype bellows was qualified for its leak tightness (less than 1 × 10{sup −9} Pam{sup 3}/s of helium) at all time during the qualification test. - Abstract: This paper presents the key manufacturing and testing processes of the prototype ITER feeder seismic bellows. The design of the double bellows was iterated with the results of analysis based on the Expansion Joint Manufacturers Association (EJMA) standard. Each inner and outer bellows was supported in dedicated molds and formed by a hydraulic pressure machine rated at 800 tons. The double bellows were constructed by welding individual collars to the end flanges. The seismic bellows was tested with cyclic pressurization of the interlayer space to 2 bars absolute pressure for 5 cycles. This was followed by 200 cycles of tensile fatigue test with 90 mm of stretching from the nominal design length. After the mechanical fatigue test, a full tensile test with 315 mm of stretching from its nominal length was conducted. Helium leak tests, with the sensitivity of the helium leak detector set to 1 × 10{sup −9} Pa m{sup 3}/s of helium, were performed at different stages of pressure and mechanical tests. The prototype bellows was qualified for its leak tightness at all time during the qualification test.

Stress analysis of a double-wall vacuum vessel for ITER

International Nuclear Information System (INIS)

Conner, D.L.; Williamson, D.E.; Nelson, B.E.

1991-01-01

The preliminary structural analyses performed in support of the design of the vacuum vessel for the International Thermonuclear Experimental Reactor (ITER) are described. A thin, double-wall, all-welded structure is the proposed design concept analyzed. The results of the static stress analysis indicate the adequacy of such a structure. The effects of the proposed high-aspect-ratio design configuration on loading and stresses are also discussed. 4 refs., 6 figs., 1 tab

Non-destructive qualification tests for ITER cryogenic axial insulating breaks

International Nuclear Information System (INIS)

Kosek, Jacek; Lopez, Roberto; Tommasini, Davide; Rodriguez-Mateos, Felix

2014-01-01

In the ITER superconducting magnets the dielectric separation between the CICC (Cable-In-Conduit Conductors) and the helium supply pipes is made through the so-called insulating breaks (IB). These devices shall provide the required dielectric insulation at a 30 kV level under different types of stresses and constraints: thermal, mechanical, dielectric and ionizing radiations. As part of the R and D program, the ITER Organization launched contracts with industrial companies aimed at the qualification of the manufacturing techniques. After reviewing the main functional aspects, this paper describes and discusses the protocol established for non-destructive qualification tests of the prototypes

Non-destructive qualification tests for ITER cryogenic axial insulating breaks

Energy Technology Data Exchange (ETDEWEB)

Kosek, Jacek [Wroclaw University of Technology, Wybrzeze Wyspianskiego 27, 50-370 Wroclaw, Poland and CERN, Geneva 23,CH-1211 (Switzerland); Lopez, Roberto; Tommasini, Davide [CERN, Geneva 23,CH-1211 (Switzerland); Rodriguez-Mateos, Felix [CERN, Geneva 23,CH-1211, Switzerland and ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul lez Durance (France)

2014-01-29

In the ITER superconducting magnets the dielectric separation between the CICC (Cable-In-Conduit Conductors) and the helium supply pipes is made through the so-called insulating breaks (IB). These devices shall provide the required dielectric insulation at a 30 kV level under different types of stresses and constraints: thermal, mechanical, dielectric and ionizing radiations. As part of the R and D program, the ITER Organization launched contracts with industrial companies aimed at the qualification of the manufacturing techniques. After reviewing the main functional aspects, this paper describes and discusses the protocol established for non-destructive qualification tests of the prototypes.

A linear iterative unfolding method

International Nuclear Information System (INIS)

László, András

2012-01-01

A frequently faced task in experimental physics is to measure the probability distribution of some quantity. Often this quantity to be measured is smeared by a non-ideal detector response or by some physical process. The procedure of removing this smearing effect from the measured distribution is called unfolding, and is a delicate problem in signal processing, due to the well-known numerical ill behavior of this task. Various methods were invented which, given some assumptions on the initial probability distribution, try to regularize the unfolding problem. Most of these methods definitely introduce bias into the estimate of the initial probability distribution. We propose a linear iterative method (motivated by the Neumann series / Landweber iteration known in functional analysis), which has the advantage that no assumptions on the initial probability distribution is needed, and the only regularization parameter is the stopping order of the iteration, which can be used to choose the best compromise between the introduced bias and the propagated statistical and systematic errors. The method is consistent: 'binwise' convergence to the initial probability distribution is proved in absence of measurement errors under a quite general condition on the response function. This condition holds for practical applications such as convolutions, calorimeter response functions, momentum reconstruction response functions based on tracking in magnetic field etc. In presence of measurement errors, explicit formulae for the propagation of the three important error terms is provided: bias error (distance from the unknown to-be-reconstructed initial distribution at a finite iteration order), statistical error, and systematic error. A trade-off between these three error terms can be used to define an optimal iteration stopping criterion, and the errors can be estimated there. We provide a numerical C library for the implementation of the method, which incorporates automatic

ITER-FEAT - outline design report. Report by the ITER Director. ITER meeting, Tokyo, January 2000

International Nuclear Information System (INIS)

2001-01-01

It is now possible to define the key elements of ITER-FEAT. This report provides the results, to date, of the joint work of the Special Working Group in the form of an Outline Design Report on the ITER-FEAT design which, subject to the views of ITER Council and of the Parties, will be the focus of further detailed design work and analysis in order to provide to the Parties a complete and fully integrated engineering design within the framework of the ITER EDA extension

Parametric analysis and operational performance of EDA-ITER

International Nuclear Information System (INIS)

Murakami, Yoshiki; Tsunematsu, Toshihide; Fujieda, Hirobumi.

1994-06-01

Confinement capability of EDA-ITER is investigated by using a 0-D model based on CDA physics design guidelines. Confinement enhancement factor (H-factor) is evaluated and required fusion power (P FUS ) for the ignition is calculated. It is found that ignition is possible in H-mode plasma (H=2) when helium accumulation (He) is 10% and P FUS ≥ 1 GW. For Rebut-Lallia scaling law, L-mode (H=1) ignition is possible when P FUS ≥ 3 GW. The required fusion power is, however, more than 4 GW even in H-mode plasmas when the helium accumulation is 20%. Therefore, it is an important future work to study how much helium accumulates in a burning plasma. Capability of steady-state mode operation is also investigated. Required current-drive power for H-mode plasma is about 140 MW when He=10% and the fusion gain Q is more than 5. If the enhanced confinement (H∼3) in high safety factor region (q∼5) can be adoptable, steady-state operation with Q>10 is possible and the required current-drive power is about 60 MW. In spite of the larger fusion power, the divertor heat load of EDA-ITER calculated by scaling models is comparable or smaller than that of CDA-ITER due to the longer connection length. Thermal instability of EDA-ITER is also investigated. The growth time is about 15 s for ITER89 power scaling law. Fusion power excursion is investigated in very preliminary way. It is found that the power rises from 1.5 GW to 3 GW in about 100 s if there is no control. Although this instability could be stabilized by beta limit or helium accumulation effect, it is an important future work since it may cause severe problem. (author)

Stress analyses of ITER toroidal field coils under fault conditions

International Nuclear Information System (INIS)

Jong, C.T.J.

1990-02-01

The International Thermonuclear Experimental Reactor (ITER) is intended as an experimental thermonuclear tokamak reactor for testing the basic physics, performance and technologies essential to future fusion reactors. The ITER design will be based on extensive new design work, supported by new physical and technological results, and on the great body of experience built up over several years from previous national and international reactor studies. Conversely, the ITER design process should provide the fusion community with valuable insights into what key areas need further development or clarification as we move forward towards practical fusion power. As part of the design process of the ITER toroidal field coils the mechanical behaviour of the magnetic system under fault conditions has to be analysed in more detail. This paper describes the work carried out to create a detailed finite element model of two toroidal field coils as well as some results of linear elastic analyses with fault conditions. The analyses have been performed with the finite element code ANSYS. (author). 5 refs.; 8 figs.; 2 tabs

Bayesian estimation of multicomponent relaxation parameters in magnetic resonance fingerprinting.

Science.gov (United States)

McGivney, Debra; Deshmane, Anagha; Jiang, Yun; Ma, Dan; Badve, Chaitra; Sloan, Andrew; Gulani, Vikas; Griswold, Mark

2018-07-01

To estimate multiple components within a single voxel in magnetic resonance fingerprinting when the number and types of tissues comprising the voxel are not known a priori. Multiple tissue components within a single voxel are potentially separable with magnetic resonance fingerprinting as a result of differences in signal evolutions of each component. The Bayesian framework for inverse problems provides a natural and flexible setting for solving this problem when the tissue composition per voxel is unknown. Assuming that only a few entries from the dictionary contribute to a mixed signal, sparsity-promoting priors can be placed upon the solution. An iterative algorithm is applied to compute the maximum a posteriori estimator of the posterior probability density to determine the magnetic resonance fingerprinting dictionary entries that contribute most significantly to mixed or pure voxels. Simulation results show that the algorithm is robust in finding the component tissues of mixed voxels. Preliminary in vivo data confirm this result, and show good agreement in voxels containing pure tissue. The Bayesian framework and algorithm shown provide accurate solutions for the partial-volume problem in magnetic resonance fingerprinting. The flexibility of the method will allow further study into different priors and hyperpriors that can be applied in the model. Magn Reson Med 80:159-170, 2018. © 2017 International Society for Magnetic Resonance in Medicine. © 2017 International Society for Magnetic Resonance in Medicine.

ITER council proceedings: 2001

International Nuclear Information System (INIS)

2001-01-01

Continuing the ITER EDA, two further ITER Council Meetings were held since the publication of ITER EDA documentation series no, 20, namely the ITER Council Meeting on 27-28 February 2001 in Toronto, and the ITER Council Meeting on 18-19 July in Vienna. That Meeting was the last one during the ITER EDA. This volume contains records of these Meetings, including: Records of decisions; List of attendees; ITER EDA status report; ITER EDA technical activities report; MAC report and advice; Final report of ITER EDA; and Press release

High voltage investigations for ITER coils

International Nuclear Information System (INIS)

Fink, S.; Fietz, W.H.

2006-01-01

The superconducting ITER magnets will be excited with high voltage during operation and fast discharge. Because the coils are complex systems the internal voltage distribution can differ to a large extent from the ideal linear voltage distribution. In case of fast excitations internal voltages between conductor and radial plate of a TF coil can be even higher than the terminal voltage of 3.5 kV to ground which appears during a fast discharge without a fault. Hence the determination of the transient voltage distribution is important for a proper insulation co-ordination and will provide a necessary basis for the verification of the individual insulation design and the choice of test voltages and waveforms. Especially the extent of internal overvoltages in case of failures, e. g. malfunction of discharge units and / or arcing is of special interest. Transient calculations for the ITER TF coil system have been performed for fast discharge and fault scenarios to define test voltages for ITER TF. The conductor and radial plate insulation of the ITER TF Model Coil were exposed at room temperature to test voltages derived from the results from these calculations. Breakdown appeared during the highest AC voltage step. A fault scenario for the TF fast discharge system is presented where one fault triggers a second fault, leading to considerable voltage stress. In addition a FEM model of Poloidal Field Coil 3 for the determination of the parameters of a detailed network model is presented in order to prepare detailed investigations of the transient voltage behaviour of the PF coils. (author)

The High Aspect Ratio Design (HARD): A candidate ITER concept with improved technology phase performance

International Nuclear Information System (INIS)

Nevins, W.M.; Perkins, L.J.; Wesley, J.C.

1992-10-01

The High Aspect Ratio Design (HARD) International Thermonuclear Experimental Reactor (ITER) concept developed by the US ITER team is an alternate to the low-aspect-ratio ITER design developed by the ITER participants during the Conceptual Design Activity (CDA). The CDA design, referred to hereafter as ITER CDA, has an aspect ratio, A, of 2.79, a toroidal magnetic field, B T , of 4.85 T, and a plasma current, I p , of 22 MA for operation with an ignited plasma. In contrast, HARD employs higher aspect ratio, A = 4.0, higher toroidal field, B T = 7.11 T, and lower plasma current, I p = 14.8 MA for ignition operation. The cross sections of the two designs are compared in. The parameters and performance of HARD and ITER CDA for inductively driven ignition operation are compared in. The HARD parameters provide the same ignition performance (ignition margin evaluated against ITER-89P confinement scaling) as ITER CDA in a device with comparable size and cost. However, the reason for advancing HARD rather than ITER CDA as the ITER design concept is not inductively driven ignition performance but HARD's significantly enhanced potential to achieve the technology testing and steady-state operation goals of the ITER objectives with non-inductive current drive

Characterizing problematic hypoglycaemia: iterative design and preliminary psychometric validation of the Hypoglycaemia Awareness Questionnaire (HypoA-Q).

Science.gov (United States)

Speight, J; Barendse, S M; Singh, H; Little, S A; Inkster, B; Frier, B M; Heller, S R; Rutter, M K; Shaw, J A M

2016-03-01

To design and conduct preliminary validation of a measure of hypoglycaemia awareness and problematic hypoglycaemia, the Hypoglycaemia Awareness Questionnaire. Exploratory and cognitive debriefing interviews were conducted with 17 adults (nine of whom were women) with Type 1 diabetes (mean ± sd age 48 ± 10 years). Questionnaire items were modified in consultation with diabetologists/psychologists. Psychometric validation was undertaken using data from 120 adults (53 women) with Type 1 diabetes (mean ± sd age 44 ± 16 years; 50% with clinically diagnosed impaired awareness of hypoglycaemia), who completed the following questionnaires: the Hypoglycaemia Awareness Questionnaire, the Gold score, the Clarke questionnaire and the Problem Areas in Diabetes questionnaire. Iterative design resulted in 33 items eliciting responses about awareness of hypoglycaemia when awake/asleep and hypoglycaemia frequency, severity and impact (healthcare utilization). Psychometric analysis identified three subscales reflecting 'impaired awareness', 'symptom level' and 'symptom frequency'. Convergent validity was indicated by strong correlations between the 'impaired awareness' subscale and existing measures of awareness: (Gold: rs =0.75, P Questionnaire has robust face and content validity; satisfactory structure; internal reliability; convergent, divergent and known groups validity. The impaired awareness subscale and other items contribute significantly to models explaining recall of severe and nocturnal hypoglycaemia. Prospective validation, including determination of a threshold to identify impaired awareness, is now warranted. © 2015 The Authors. Diabetic Medicine © 2015 Diabetes UK.

Poloidal field system for the ITER hard design option

International Nuclear Information System (INIS)

Schultz, J.H.; Pillsbury, R.D.

1992-01-01

This paper reports on ITER, the International Thermonuclear Experimental Reactor, a collaborative design by the US, EC, Japan, and the USSR of a tokamak fusion reactor that will demonstrate the physics and test the technology needed for commercial fusion reactors. In 1990, the ITER team completed a Conceptual Design Activity (CDA) in which a candidate design was shown to meet the specified goals of the ITER activity at a conceptual level. The four parties have agreed to an Engineering Design Activity (EDA) that includes the necessary additional design and analysis, along with the R and D needed to construct ITER with confidence. The CDA design includes a toroidal field (TF) magnet system that provides the main containment field and a poloidal field (PF) system used to control plasma current and position. The PF system is also used as transformer primary to induce and sustain current in the plasma. Since the volt-seconds available for full-current plasma burn are less than 10% of the total available volt-seconds from the PF system, an area of concern in the CDA design is that unfavorable plasma conditions could compromise the ability of the physics base case design to achieve long pulse burns. A High Aspect Ratio Design (HARD) was conceived as an alternative design option with a much larger bore in the central solenoid to enhance ITER's capabilities for long-burn operation

Fusion Plasma Physics and ITER - An Introduction (1/4)

CERN Multimedia

CERN. Geneva

2011-01-01

In November 2006, ministers representing the world’s major fusion research communities signed the agreement formally establishing the international project ITER. Sited at Cadarache in France, the project involves China, the European Union (including Switzerland), India, Japan, the Russian Federation, South Korea and the United States. ITER is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. The ITER tokamak is designed to study the “burning plasma” regime in deuterium-tritium (D-T) plasmas by achieving a fusion amplification factor, Q (the ratio of fusion output power to plasma heating input power), of 10 for several hundreds of seconds with a nominal fusion power output of 500MW. It is also intended to allow the study of steady-state plasma operation at Q≥5 by me...

Investigation of cryogenic irradiation influence on mechanical and physical properties of ITER magnetic system insulation materials

International Nuclear Information System (INIS)

Kozlov, A.V.; Scherbacov, E.N.; Dudchenko, N.A.; Shihalev, V.S.; Bedin, V.V.; Paltusov, N.A.; Korsunskiy, V.E.

1998-01-01

A set of methods of cryogenic irradiation influence test on mechanical and physical properties of insulation of ITER magnetic system are presented in this paper. Investigations are carried out without intermediate warming up of samples. A Russian insulating composite material was irradiated in the IVV-2M reactor. The ratio of energy absorbed by insulation materials from neutron irradiation to that from gamma irradiation can be varied from ∝(25:75)% to ∝(50:50)% in the reactor. The test results on the thermal expansion, thermal conductivity and gas evolution of the above material are presented. It was shown, that cryogenic irradiation up to the fluence ∝2 x 10 22 n/m 2 (E ≥ 0.1 MeV) leads to 0.27% linear size changes along layers of fiber-glass, the thermal conductivity coefficient is decreased on 15% at 100 k in perpendicular direction to fiber-glass plane, and thermal coefficient of linear expansion (TCLE) has anomalous temperature dependence. (orig.)

The First Benchmarking of ITER BR Nb3Sn Strand of CNDA

International Nuclear Information System (INIS)

Long Feng; Liu Fang; Wu Yu; Ni Zhipeng

2012-01-01

According to the International Thermonuclear Experimental Reactor (ITER) Procurement Arrangement (PA) of Cable-In-Conduit Conductor (CICC) unit lengths for the Toroidal Field (TF) and Poloidal Field (PF) magnet systems of ITER, at the start of process qualification, the Domestic Agency (DA) shall be required to conduct a benchmarking of the room and low temperature acceptance tests carried out at the Strand Suppliers and/or at its Reference Laboratories designated by the ITER Organization (IO). The first benchmarking was carried out successfully in 2009. Nineteen participants from six DAs (China, European Union, Japan, South Korea, Russia, and the United States) participated in the first benchmarking. Bronze-route (BR) Nb 3 Sn strand and samples prepared by the ITER reference lab (CERN) were sent out to each participant by CERN. In this paper, the test facility and test results of the first benchmarking by the Chinese DA (CNDA) are presented.

Final report of the ITER EDA. Final report of the ITER Engineering Design Activities. Prepared by the ITER Council

International Nuclear Information System (INIS)

2001-01-01

This is the Final Report by the ITER Council on work carried out by ITER participating countries on cooperation in the Engineering Design Activities (EDA) for the ITER. In this report the main ITER EDA technical objectives, the scope of ITER EDA, its organization and resources, engineering design of ITER tokamak and its main parameters are presented. This Report also includes safety and environmental assessments, site requirements and proposed schedule and estimates of manpower and cost as well as proposals on approaches to joint implementation of the project

Planning for U.S. Fusion Community Participation in the ITER Program

International Nuclear Information System (INIS)

Baker, Charles; Berk, Herbert; Greenwald, Martin; Mauel, Michael E.; Najmabadi, Farrokh; Nevins, William M.; Stambaugh, Ronald; Synakowski, Edmund; Batchelor, Donald B.; Fonck, Raymond; Hawryluk, Richard J.; Meade, Dale M.; Neilson, George H.; Parker, Ronald; Strait, Ted

2006-01-01

A central step in the mission of the U.S. Fusion Energy Sciences program is the creation and study of a fusion-powered 'star on earth', where the same energy source that drives the sun and other stars is reproduced and controlled for sustained periods in the laboratory. This ''star'' is formed by an ionized gas, or plasma, heated to fusion temperatures in a magnetic confinement device known as a tokamak, which is the most advanced magnetic fusion concept. The ITER tokamak is designed to be the premier scientific tool for exploring and testing expectations for plasma behavior in the fusion burning plasma regime, wherein the fusion process itself provides the dominant heat source to sustain the plasma temperature. It will provide the scientific basis and control tools needed to move toward the fusion energy goal. The ITER project confronts the grand challenge of creating and understanding a burning plasma for the first time. The distinguishing characteristic of a burning plasma is the tight coupling between the fusion heating, the resulting energetic particles, and the confinement and stability properties of the plasma. Achieving this strongly coupled burning state requires resolving complex physics issues and integrating challenging technologies. A clear and comprehensive scientific understanding of the burning plasma state is needed to confidently extrapolate plasma behavior and related technology beyond ITER to a fusion power plant. Developing this predictive understanding is the overarching goal of the U.S. Fusion Energy Sciences program. The burning plasma research program in the U.S. is being organized to maximize the scientific benefits of U.S. participation in the international ITER experiment. It is expected that much of the research pursued on ITER will be based on the scientific merit of proposed activities, and it will be necessary to maintain strong fusion research capabilities in the U.S. to successfully contribute to the

Planning for U.S. Fusion Community Participation in the ITER Program

Energy Technology Data Exchange (ETDEWEB)

Baker, Charles [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Berk, Herbert [Univ. of Texas, Austin, TX (United States); Greenwald, Martin [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mauel, Michael E. [Columbia Univ., New York, NY (United States); Najmabadi, Farrokh [Univ. of California, San Diego, CA (United States); Nevins, William M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Stambaugh, Ronald [General Atomics, La Jolla, CA (United States); Synakowski, Edmund [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Batchelor, Donald B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fonck, Raymond [Univ. of Wisconsin, Madison, WI (United States); Hawryluk, Richard J. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Meade, Dale M. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Neilson, George H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Parker, Ronald [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Strait, Ted [General Atomics, La Jolla, CA (United States)

2006-06-07

A central step in the mission of the U.S. Fusion Energy Sciences program is the creation and study of a fusion-powered "star on earth", where the same energy source that drives the sun and other stars is reproduced and controlled for sustained periods in the laboratory. This “star” is formed by an ionized gas, or plasma, heated to fusion temperatures in a magnetic confinement device known as a tokamak, which is the most advanced magnetic fusion concept. The ITER tokamak is designed to be the premier scientific tool for exploring and testing expectations for plasma behavior in the fusion burning plasma regime, wherein the fusion process itself provides the dominant heat source to sustain the plasma temperature. It will provide the scientific basis and control tools needed to move toward the fusion energy goal. The ITER project confronts the grand challenge of creating and understanding a burning plasma for the first time. The distinguishing characteristic of a burning plasma is the tight coupling between the fusion heating, the resulting energetic particles, and the confinement and stability properties of the plasma. Achieving this strongly coupled burning state requires resolving complex physics issues and integrating challenging technologies. A clear and comprehensive scientific understanding of the burning plasma state is needed to confidently extrapolate plasma behavior and related technology beyond ITER to a fusion power plant. Developing this predictive understanding is the overarching goal of the U.S. Fusion Energy Sciences program. The burning plasma research program in the U.S. is being organized to maximize the scientific benefits of U.S. participation in the international ITER experiment. It is expected that much of the research pursued on ITER will be based on the scientific merit of proposed activities, and it will be necessary to maintain strong fusion research capabilities in the U.S. to successfully contribute to the success of ITER and optimize

Preliminary magnetostratigraphy and environmental magnetism of the Lower Cretaceous from the Italian Dolomites

Science.gov (United States)

Savian, J. F.; Jovane, L.; Florindo, F.; Lukeneder, A.

2011-12-01

The Lower Cretaceous (~146 to 100 Ma) represents an enigmatic time interval for paleoclimatic, paleogeography and paleomagnetic evolution of the Earth's history. The climatic changes include global oceanic anoxic events (OAEs), biotic changes, global excursions of carbon and strontium isotopes, rises in eustatic sea level and paleotemperature. Paleoceanography was marked by a rapid rate of ocean spreading in the Atlantic. The opening of the Atlantic Ocean was wide enough to allow significant circulation of masses of waters across the equator. This period is furthermore important for the oceanographic events occurring at the base of the Aptian (Selli Level). This period also present one of the most intriguing geomagnetic events: the long normal Cretaceous superchron, lasted for almost 40 million years. We study here the lower Cretaceous deposits of the Puez section in the Dolomites (northern Italy) which represents a continuous section during this period. The samples collected represent marine sedimentary materials of the Biancone and Puez formations. The Puez section consists essentially of green-grey to red limestones and calcareous marls. We present preliminary results of integrated magnetostratigraphic analysis, including a detailed lithostratigraphy and environmental magnetism. We recognize magnetic behavior that are relative to normal polarity (the normal Cretaceous superchron), with a short reverse interval that might represent the M-1r event. We also recognize a series of normal and reverse polarities (below the normal Cretaceous superchron) which can be referred to the magnetozones M1/M5. The environmental magnetic data consists of magnetic susceptibility (χ), natural remanent magnetization (NRM), anhysteretic remanent magnetization (ARM), isothermal remanent magnetization (IRM) at 900 mT and backfield isothermal remanent magnetization (BIRM) at 100 mT and 300 mT. Derived parameters, such as S-ratio (S300=BIRM300/IRM900) and hard isothermal remanent

Preliminary design of bellows for the DNB beam source by EJMA and FE linear analysis

International Nuclear Information System (INIS)

Trapasiya, Shobhit; Muvvala, Venkata Nagaraju; Rambilas, P.; Gangadharan, Roopesh; Rotti, Chandramouli; Chakraborty, Arun Kumar; Sharma, Dheeraj Kumar

2015-01-01

In piping system, U-shaped Bellows are widely used among flexible elements. In general, bellows are typically design for Fatigue behavior according to the EJMA standard based on empirically generated fatigue curves. The present work proposes a methodology in the design of bellows by design by analyses and validates its design by EJMA standard. A linear FE approach is chosen to in line with the EJMA standard. The proposed methodology is benchmarked with the available literatures. The same practice is implemented in the preliminary design of a U-shaped bellows in the water line circuits of DNB beam source. DNB Beam Source is a negative ion source-based neutral beam generator for ITER operates at 100KV. The beam divergence (intrinsic) and magnetic fields from ITER torus causes deflection of beams. This calls for beam optic alignment, which are assured by BS Movement mechanism system. To accomplish the above movement requirements, bellows, which is a stringent of its kind (± 22 mm axial, ± 45 mm lateral within 400mm available space with single ply), is designed between the beam source and possible rigid interface-cooling lines coming from HVB. The paper describes right from conceptual stage to preliminary design. Optimization tools are adopted in the selecting bellow dimensions using MATLAB. At the end a coordinated approach between FE based assessment (in ANSYS) and widely applied code, EJMA is implemented for the validation of design and found FE approach is a very conservative than later in the present case. (author)

Series Supply of Cryogenic Venturi Flowmeters for the ITER Project

International Nuclear Information System (INIS)

André, J; Poncet, J M; Ercolani, E; Clayton, N; Journeaux, J Y

2015-01-01

In the framework of the ITER project, the CEA-SBT has been contracted to supply 277 venturi tube flowmeters to measure the distribution of helium in the superconducting magnets of the ITER tokamak. Six sizes of venturi tube have been designed so as to span a measurable helium flowrate range from 0.1 g/s to 400g/s. They operate, in nominal conditions, either at 4K or at 300K, and in a nuclear and magnetic environment. Due to the cryogenic conditions and the large number of venturi tubes to be supplied, an individual calibration of each venturi tube would be too expensive and time consuming. Studies have been performed to produce a design which will offer high repeatability in manufacture, reduce the geometrical uncertainties and improve the final helium flowrate measurement accuracy. On the instrumentation side, technologies for differential and absolute pressure transducers able to operate in applied magnetic fields need to be identified and validated. The complete helium mass flow measurement chain will be qualified in four test benches: - A helium loop at room temperature to insure the qualification of a statistically relevant number of venturi tubes operating at 300K.- A supercritical helium loop for the qualification of venturi tubes operating at cryogenic temperature (a modification to the HELIOS test bench). - A dedicated vacuum vessel to check the helium leak tightness of all the venturi tubes. - A magnetic test bench to qualify different technologies of pressure transducer in applied magnetic fields up to 100mT. (paper)

iterClust: a statistical framework for iterative clustering analysis.

Science.gov (United States)

Ding, Hongxu; Wang, Wanxin; Califano, Andrea

2018-03-22

In a scenario where populations A, B1 and B2 (subpopulations of B) exist, pronounced differences between A and B may mask subtle differences between B1 and B2. Here we present iterClust, an iterative clustering framework, which can separate more pronounced differences (e.g. A and B) in starting iterations, followed by relatively subtle differences (e.g. B1 and B2), providing a comprehensive clustering trajectory. iterClust is implemented as a Bioconductor R package. andrea.califano@columbia.edu, hd2326@columbia.edu. Supplementary information is available at Bioinformatics online.

A numerical model for the simulation of quench in the ITER magnets

International Nuclear Information System (INIS)

Bottura, L.

1996-01-01

A computational model describing the initiation and evolution of normal zones in the cable-in-conduit superconductors designed for the international thermonuclear experimental reactor (ITER) is presented. Because of the particular geometry of the ITER cables, the model treats separately the helium momenta in the two cooling channels and the temperatures of the cable constituents. The numerical implementation of the model is discussed in conjunction with the selection of a well-suited solution algorithm. In particular, the solution procedure chosen is based on an implicit upwind finite element technique with adaptive time step and mesh size adjustment possibilities. The time step and mesh adaption procedures are described. Examples of application of the model are also reported. 39 refs., 6 figs., 2 tabs

ITER-FEAT vacuum pumping and fuelling R and D programmes

International Nuclear Information System (INIS)

Murdoch, D.K.; Antipenkov, A.; Ladd, P.; Boissin, J.-C.; Day, C.; Haas, H.; Mack, A.; Pimanikhin, S.; Saksagansky, G.; Viniar, I.

2001-01-01

The design of the ITER-FEAT vacuum pumping and fuelling systems is supported by two key R and D programs, the first directed towards the development of a steady state tritium compatible pellet injector, and the second towards the development of a supercritical helium cooled cryogenic pump for torus exhaust. While the pellet injector programme for ITER-FEAT is new, that for the cryopump has evolved from a programme that originally supported the 1998 ITER design. As the plasma exhaust parameters have remained essentially unchanged between these two machines, the R and D conducted to date remains valid. Initial test results on the prototype injector, TPI-1, which included continuous injection of 3 mm hydrogen pellets at 500 m/s and at 1 to 2 Hz for periods up to, are reported. A model of the cryopump has now been installed in a new dedicated test bed at the Karlsruhe Research Centre where acceptance tests have been completed and preliminary results from pumping tests obtained. An extensive test campaign to fully characterise pump performance and identify any mechanical details which require modification has started. (author)

Conceptual design and related R and D on ITER mechanical based primary pumping system

International Nuclear Information System (INIS)

Tanzawa, Sadamitsu; Hiroki, Seiji; Abe, Tetsuya; Shimizu, Katsusuke; Inoue, Masahiko; Watanabe, Mitsunori; Iguchi, Masashi; Sugimoto, Tomoko; Inohara, Takashi; Nakamura, Jun-ichi

2008-12-01

The primary vacuum pumping system of the International Thermonuclear Experimental Reactor (ITER) exhausts a helium (He) ash resulting from the DT-burn with excess DT fueling gas, as well as performing a variety of functions such as pump-down, leak testing and wall conditioning. A mechanical based vacuum pumping system has some merits of a continuous pumping, a much lower tritium inventory, a lower operational cost and easy maintenance, comparing with a cryopump system, although demerits of an indispensable magnetic shield and insufficient performance for hydrogen (H 2 ) pumping is are well recognized. To overcome the demerits, we newly fabricated and tested a helical grooved pump (HGP) unit suitable for H 2 pumping at the ITER divertor pressure of 0.1-10 Pa. Through this R and D, we successfully established many design and manufacturing databases of large HGP units for the lightweight gas pumping. Based on the databases, we conceptually designed the ITER vacuum pumping system mainly comprising the HGP with an optimal pump unit layout and a magnetic shield. We also designed conceptually the reduced cost (RC)-ITER pumping system, where a compound molecular pump combining turbine bladed rotors and helical grooved ones was mainly used. The ITER mechanical based primary pumping system proposed has eventually been a back-up solution, whereas a cryopump based one was formally selected to the ITER for construction. The mechanical pumps are increasingly used in many areas with well sophisticated performance, so we believe that fusion reactors of subsequent prototype ones will select the mechanical based pumping system due to primarily a high operational reliability and a cost melt. (author)

ITER-EDA physics design requirements and plasma performance assessments

International Nuclear Information System (INIS)

Uckan, N.A.; Galambos, J.; Wesley, J.; Boucher, D.; Perkins, F.; Post, D.; Putvinski, S.

1996-01-01

Physics design guidelines, plasma performance estimates, and sensitivity of performance to changes in physics assumptions are presented for the ITER-EDA Interim Design. The overall ITER device parameters have been derived from the performance goals using physics guidelines based on the physics R ampersand D results. The ITER-EDA design has a single-null divertor configuration (divertor at the bottom) with a nominal plasma current of 21 MA, magnetic field of 5.68 T, major and minor radius of 8.14 m and 2.8 m, and a plasma elongation (at the 95% flux surface) of ∼1.6 that produces a nominal fusion power of ∼1.5 GW for an ignited burn pulse length of ≥1000 s. The assessments have shown that ignition at 1.5 GW of fusion power can be sustained in ITER for 1000 s given present extrapolations of H-mode confinement (τ E = 0.85 x τ ITER93H ), helium exhaust (τ* He /τ E = 10), representative plasma impurities (n Be /n e = 2%), and beta limit [β N = β(%)/(I/aB) ≤ 2.5]. The provision of 100 MW of auxiliary power, necessary to access to H-mode during the approach to ignition, provides for the possibility of driven burn operations at Q = 15. This enables ITER to fulfill its mission of fusion power (∼ 1--1.5 GW) and fluence (∼1 MWa/m 2 ) goals if confinement, impurity levels, or operational (density, beta) limits prove to be less favorable than present projections. The power threshold for H-L transition, confinement uncertainties, and operational limits (Greenwald density limit and beta limit) are potential performance limiting issues. Improvement of the helium exhaust (τ* He /τ E ≤ 5) and potential operation in reverse-shear mode significantly improve ITER performance

Research and development needs for ITER engineering design

International Nuclear Information System (INIS)

Flanagan, C.; Alikaev, V.; Baker, C.

1991-01-01

In the series of documents that summarize the results of the Conceptual Design Activities (CDA) for the International Thermonuclear Experimental Reactor (ITER), this document describes the research and development (R and D) plans for 1991 - 1995. Part A describes the physics R and D, part B the technology R and D. The Physics R and D needs are presented in terms of task descriptions of an ITER-related R and D programme for 1991/1992 and beyond, while diagnostics R and D needs, although covered in Appendix A, are described in Part B. In Chapter II of Part A, ''ITER-related Physics R and D Needs for 91/92 and Beyond'', the following tasks are described as most crucial: (1) demonstration that (i) operation with a cold divertor plasma is possible, (ii) the peak heat flux onto the divertor plate can be kept below about 10 MW per square meter, (iii) and helium exhaust conditions allow a fractional burnup of about 3 percent or more; (2) a characterisation of disruptions that allows to specify their consequences for the plasma-facing-components, and that provides evidence that the number of disruptions expected allows acceptable plasma-facing-component lifetimes; (3) demonstration that steady-state operation in an enhanced-confinement regime and satisfactory plasma purity is possible, and provision of energy confinement scaling allowing the prediction of ITER performance; and (4) ensurance that the presence of a fast ion population does not jeopardize plasma performance in ITER. Part B, ''ITER Technology Research and Development Needs'', describes planning R and D for magnets, containment structure, assembly and maintenance, current drive and heating, plasma facing components, blanket, fuel cycle, structural materials, and diagnostics. A table of key milestones for Technology R and D is included, as well as cost estimates. Figs and tabs

Minimum dimension of an ITER like Tokamak with a given Q

Energy Technology Data Exchange (ETDEWEB)

Johner, J

2004-07-01

The minimum dimension of an ITER like tokamak with a given amplification factor Q is calculated for two values of the maximum magnetic field in the superconducting toroidal field coils. For ITERH-98P(y,2) scaling of the energy confinement time, it is shown that for a sufficiently large tokamak, the maximum Q is obtained for the operating point situated both at the maximum density and at the minimum margin with respect to the H-L transition. We have shown that increasing the maximum magnetic field in the toroidal field coils from the present 11.8 T to 16 T would result in a strong reduction of the machine size but has practically no effect on the fusion power. Values obtained for {beta}{sub N} are found to be below 2. Peak fluxes on the divertor plates with an ITER like divertor and a multi-machine expression for the power radiated in the plasma mantle, are below 10 MW/m{sup 2}.

ITER, On the way to call for the license

International Nuclear Information System (INIS)

Alejaidre, C.; Girard, J.P.

2007-01-01

Full text of publication follows: All the seven ITER Parties have now ratified the lTER Agreement, which was signed in Paris in November last year. On October 24, 2007 the ITER Agreement enters officially into force. The licensing procedure to start construction shall be launched within the coming months in order to start construction at the beginning of 2009. The Host Party, Europe and the Host State, France have already participated together with the ITER Organization (IO) to a public debate to present the project to the neighboring populations (January to May 2006) and other administrative procedures to prepare the site and the roads for heavy loads. According to host state law on licensing of nuclear facilities, IO will send a set of administrative documents including a safety assessment (called Preliminary Safety Report) and public hearing files including impact studies and risk analysis. The review and public hearing are expected mid-08 to be consistent with the start of nuclear building construction on site. The latest update of the design took into account of inputs from: -An internal design review managed through eight specialist groups including safety and licensing, - Host State regulations (technical and QA), which should be fulfilled as quoted in the ITER agreement, - Site adaptation including land configuration, environmental conditions and external hazards. The ITER designers, in close contact with the Participant Teams, are proceeding with the adaptation of the generic design to comply with these new inputs. The codes and standards for all equipment are also under revision in order to fit with the expected requirements, taking into account the procurement sharing agreement and the French regulations. In the earlier stage of ITER design fusion took benefit from fission knowledge, in the present stage fusion is taking the lead as can be demonstrated by the update of the European mechanical code RCC-MR, now ready and updated to the vacuum vessel design

Neutral particle beam alternative concept for ITER

International Nuclear Information System (INIS)

Sedgley, D.; Brook, J.; Luzzi, T.; Deutsch, L.

1989-01-01

An analysis of an ITER neutral particle beam system is presented. The analysis covers the neutralizer, ion dumps, pumping, and geometric aspects. The US beam concept for ITER consists of three or four clusters of beamlines delivering approximately 80 MW total of 1.6-MeV deuterium to three or four reactor ports. Each cluster has three self-contained beamlines featuring plasma neutralizers and electrostatic ion dumps. In this study, each of the beamlines has two source assemblies with separate gas neutralizers and magnetic ion dumps. Deuterium is injected into the gas neutralizers by a separate system. Saddle-shaped copper coils augment the tokamak poloidal field to turn the charged particles into the ion dumps. The gas flow from the source, neutralizer, and ion dump is pumped by regenerable cryopanels. The effect of the port between the TF coils and the beam injection angle on the plasma footprint was studied

The Tritium Systems Test Assembly applicability to ITER

International Nuclear Information System (INIS)

Anderson, J.L.

1988-01-01

The Tritium Systems Test Assembly (TSTA), is operated by the Los Alamos National Laboratory (LANL) under the sponsorship of the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). The objectives of the TSTA project are to develop, demonstrate, and evaluate the exhaust gas processing and tritium related safety systems for the magnetic fusion energy program. The applicability of these processes for the ITER Tokamak is discussed

Preliminary analysis of hydrogen risk caused by dust in ITER

International Nuclear Information System (INIS)

Cheng Kun; Tong Lili; Cao Xuewu

2012-01-01

A lot of dust will be generated during ITER operation,and hydrogen will be produced by the interaction of hot dust with water in the case of coolant ingress accident. The accumulated hydrogen will bring risk of combustion and explosion,which will damage the device. CFD method has been used to analyze the produced hydrogen in 'wet bypass' scenario, and come to the results that hydrogen will burn and explode at the beginning of the accident, different hydrogen risk will be brought by different coolant leakage, and hydrogen risk will be inert if the leakage is massive.Injecting CO 2 to inert the vacuum vessel has also been discussed, the risk of hydrogen will be suppressed by injecting CO 2 with a large rate at the beginning of accident. (authors)

Study and optimization of magnetized ICRF discharges for tokamak wall conditioning and assessment of the applicability to ITER

International Nuclear Information System (INIS)

Wauters, T.

2011-11-01

This work is devoted to the study and optimization of the Ion Cyclotron Wall Conditioning (ICWC) technique. ICWC, operated in presence of the toroidal magnetic field, makes use of four main tokamak systems: the ICRF antennas to initiate and sustain the conditioning discharge, the gas injection valves to provide the discharge gas, the machine pumps to remove the wall desorbed particles, and the poloidal magnetic field system to optimize the discharge homogeneity. Additionally neutral gas and plasma diagnostics are required to monitor the discharge and the conditioning efficiency. In chapter 2 a general overview on ICWC is given. Chapter 3 treats the ICRF discharge homogeneity and the confinement properties of the employed magnetic field. In the first part we will discuss experimental facts on plasma homogeneity, and how experimental optimization led to its improvement. In the second part of the chapter the confinement properties of a partially ionized plasma in a toroidal magnetic field configuration with additional small vertical component are discussed. Chapter 4 gives an overview of experimental results on the efficiency of ICWC, obtained on TORE SUPRA, TEXTOR, JET and ASDEX Upgrade. In chapter 5 a 0D kinetic description of hydrogen-helium RF plasmas is outlined. The model, describing the evolution of ICRF plasmas from discharge initiation to the (quasi) steady state plasma stage, is developed to obtain insight on ICRF plasma parameters, particle fluxes to the walls and the main collisional processes. Chapter 6 presents a minimum structure for a 0D reservoir model of the wall to investigate in deeper detail the ICWC plasma wall interaction during isotopic exchange experiments. The hypothesis used to build up the wall model is that the same model structure should be able to describe the wall behavior during normal plasmas and conditioning procedures. Chapter 7 extrapolates the results to the envisaged application of ICWC on ITER

ITER council proceedings: 2000

International Nuclear Information System (INIS)

2001-01-01

No ITER Council Meetings were held during 2000. However, two ITER EDA Meetings were held, one in Tokyo, January 19-20, and one in Moscow, June 29-30. The parties participating in these meetings were those that partake in the extended ITER EDA, namely the EU, the Russian Federation, and Japan. This document contains, a/o, the records of these meetings, the list of attendees, the agenda, the ITER EDA Status Reports issued during these meetings, the TAC (Technical Advisory Committee) reports and recommendations, the MAC Reports and Advice (also for the July 1999 Meeting), the ITER-FEAT Outline Design Report, the TAC Reports and Recommendations both meetings), Site requirements and Site Design Assumptions, the Tentative Sequence of technical Activities 2000-2001, Report of the ITER SWG-P2 on Joint Implementation of ITER, EU/ITER Canada Proposal for New ITER Identification

Magnetization and loss measurements on Nb3Sn and NbTi strands for ITER and LHC

International Nuclear Information System (INIS)

Foitl, M.

2001-05-01

this work, magnetization and loss measurements on 20 different strand samples which were taken from different billets supplied by two manufacturers of NbTi strands are reported, which with the use of an Integrating Coil Magnetometer as well as micro Hall sensors magnetization loops of single strands were studied. Variations in the strand magnetization could be either correlated with systematical irregularities in the strand characteristics (e.g. filament diameter) or with variations of the strand's critical current density jC. The second part of this work is related to the annealing process of superconducting magnets manufactured from internal tin Nb3Sn strands to be used for the magnetic confinement of fusion plasma. After the successful generation of plasma burning pulses of several seconds duration (Joint European Torus, JET), magnetic fusion energy research has reached a point where a tokamak burning plasma facility, in which the thermonuclear heating balances transport and radiation losses for periods of 500 s or longer, can be seriously contemplated as a next step. Achieving this goal would be a major step forward, both in science and technology, towards the ultimate goal of magnetic fusion generation of electrical power. Therefore the main objectives of the International Thermonuclear Experimental Reactor (ITER) will be the demonstration of the scientific and technological feasibility of fusion energy on a scale close to that of an eventual thermonuclear power reactor. The magnitude of the magnetic field (∼ 11.8 T) needed to confine stable a plasma of sufficient pressure to generate ∼ 0.5 GW of fusion power is comparable to the limiting magnetic fields that a toroidal superconducting magnet system can produce. In order to achieve the best magnet performance possible, the toroidal field (TF) coils made of superconducting Nb3Sn cable-in-conduit type conductors have to be optimized with respect to the maximum transport current and transient field losses

Perl Modules for Constructing Iterators

Science.gov (United States)

Tilmes, Curt

2009-01-01

The Iterator Perl Module provides a general-purpose framework for constructing iterator objects within Perl, and a standard API for interacting with those objects. Iterators are an object-oriented design pattern where a description of a series of values is used in a constructor. Subsequent queries can request values in that series. These Perl modules build on the standard Iterator framework and provide iterators for some other types of values. Iterator::DateTime constructs iterators from DateTime objects or Date::Parse descriptions and ICal/RFC 2445 style re-currence descriptions. It supports a variety of input parameters, including a start to the sequence, an end to the sequence, an Ical/RFC 2445 recurrence describing the frequency of the values in the series, and a format description that can refine the presentation manner of the DateTime. Iterator::String constructs iterators from string representations. This module is useful in contexts where the API consists of supplying a string and getting back an iterator where the specific iteration desired is opaque to the caller. It is of particular value to the Iterator::Hash module which provides nested iterations. Iterator::Hash constructs iterators from Perl hashes that can include multiple iterators. The constructed iterators will return all the permutations of the iterations of the hash by nested iteration of embedded iterators. A hash simply includes a set of keys mapped to values. It is a very common data structure used throughout Perl programming. The Iterator:: Hash module allows a hash to include strings defining iterators (parsed and dispatched with Iterator::String) that are used to construct an overall series of hash values.

ITER overview

International Nuclear Information System (INIS)

Shimomura, Y.; Aymar, R.; Chuyanov, V.; Huguet, M.; Parker, R.R.

2001-01-01

This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

ITER Overview

International Nuclear Information System (INIS)

Shimomura, Y.; Aymar, R.; Chuyanov, V.; Huguet, M.; Parker, R.

1999-01-01

This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

Sparse BLIP: BLind Iterative Parallel imaging reconstruction using compressed sensing.

Science.gov (United States)

She, Huajun; Chen, Rong-Rong; Liang, Dong; DiBella, Edward V R; Ying, Leslie

2014-02-01

To develop a sensitivity-based parallel imaging reconstruction method to reconstruct iteratively both the coil sensitivities and MR image simultaneously based on their prior information. Parallel magnetic resonance imaging reconstruction problem can be formulated as a multichannel sampling problem where solutions are sought analytically. However, the channel functions given by the coil sensitivities in parallel imaging are not known exactly and the estimation error usually leads to artifacts. In this study, we propose a new reconstruction algorithm, termed Sparse BLind Iterative Parallel, for blind iterative parallel imaging reconstruction using compressed sensing. The proposed algorithm reconstructs both the sensitivity functions and the image simultaneously from undersampled data. It enforces the sparseness constraint in the image as done in compressed sensing, but is different from compressed sensing in that the sensing matrix is unknown and additional constraint is enforced on the sensitivities as well. Both phantom and in vivo imaging experiments were carried out with retrospective undersampling to evaluate the performance of the proposed method. Experiments show improvement in Sparse BLind Iterative Parallel reconstruction when compared with Sparse SENSE, JSENSE, IRGN-TV, and L1-SPIRiT reconstructions with the same number of measurements. The proposed Sparse BLind Iterative Parallel algorithm reduces the reconstruction errors when compared to the state-of-the-art parallel imaging methods. Copyright © 2013 Wiley Periodicals, Inc.

Development and Reliability of a Preliminary Foot Osteoarthritis Magnetic Resonance Imaging Score.

Science.gov (United States)

Halstead, Jill; Martín-Hervás, Carmen; Hensor, Elizabeth M A; McGonagle, Dennis; Keenan, Anne-Maree; Redmond, Anthony C; Conaghan, Philip G

2017-08-01

Foot osteoarthritis (OA) is very common but underinvestigated musculoskeletal condition and there is little consensus as to common magnetic resonance imaging (MRI) features. The aim of this study was to develop a preliminary foot OA MRI score (FOAMRIS) and evaluate its reliability. This preliminary semiquantitative score included the hindfoot, midfoot, and metatarsophalangeal joints. Joints were scored for joint space narrowing (JSN; 0-3), osteophytes (0-3), joint effusion/synovitis, and bone cysts (present/absent). Erosions and bone marrow lesions (BML) were scored (0-3) and BML were evaluated adjacent to entheses and at sub-tendon sites (present/absent). Additionally, tenosynovitis (0-3) and midfoot ligament pathology (present/absent) were scored. Reliability was evaluated in 15 people with foot pain and MRI-detected OA using 3.0T MRI multi-sequence protocols, and assessed using ICC as an overall score and per anatomical site. Intrareader agreement (ICC) was generally good to excellent across the foot in joint features (JSN 0.90, osteophytes 0.90, effusion/synovitis 0.46, cysts 0.87), bone features (BML 0.83, erosion 0.66, BML entheses 0.66, BML sub-tendon 0.60) and soft tissue features (tenosynovitis 0.83, ligaments 0.77). Interreader agreement was lower for joint features (JSN 0.43, osteophytes 0.27, effusion/synovitis 0.02, cysts 0.48), bone features (BML 0.68, erosion 0.00, BML entheses 0.34, BML sub-tendon 0.13), and soft tissue features (tenosynovitis 0.35, ligaments 0.33). This preliminary FOAMRIS demonstrated good intrareader reliability and fair interreader reliability when assessing the total feature scores. Further development is required in cohorts with a range of pathologies and to assess the psychometric measurement properties.

ITER Council proceedings: 1993

International Nuclear Information System (INIS)

1994-01-01

Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ''ITER EDA Agreement and Protocol 2'' (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tasks; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings

ITER council proceedings: 1993

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-31

Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ``ITER EDA Agreement and Protocol 2`` (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tastes; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings.

Some aspects of the design of the ITER NBI Active Correction and Compensation Coils

Energy Technology Data Exchange (ETDEWEB)

Alonso, Javier, E-mail: javier.alonso@ciemat.es [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); Barrera, Germán; Cabrera, Santiago; Rincón, Esther; Ríos, Luis; Soleto, Alfonso [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); El-Ouazzani, Anass; Graceffa, Joseph; Shah, Darshan; Urbani, Marc [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3 – 07/08, 08019 Barcelona (Spain)

2015-10-15

Highlights: • Water cooled coil design. • Magnetic shielding of the plasma heating Neutral Beam Injection System. • Active coils for magnetic field compensation. - Abstract: The neutral beam system for ITER consists of two heating and current drive injectors plus a diagnostic neutral beam injector. The proposed physical plant layout allows for a possible third heating injector to be installed later. For correct operation of the beam source, and to avoid deflections of the charged fraction of the beam, the magnetic field along the beam path must be very low. To minimize the stray ITER field in critical areas (ion source, acceleration grids, neutralizer, residual ion dump), a Magnetic Field Reduction System will envelop the beam vessels and the high voltage transmission lines to ion source. This whole system comprises the Passive Magnetic Shield, a set of thick steel plates, and the Active Correction and Compensation Coils, a set of coils carrying currents which depend on the tokamak stray field. This paper describes the status of the coil design, terminals and support structures, as well as a description of the calculations carried out. Most coils are suitable for removal from their final position to be replaced in case of a fault. Conclusions of the chosen design highlight the strategy for the system feasibility.

A new theoretical probe for the magnetic force microscope

Energy Technology Data Exchange (ETDEWEB)

Windmill, J.F.C. E-mail: jwindmill@plymouth.ac.uk; Clegg, W.W.; Jenkins, D.F.L.; Davey, P.J

2001-05-01

The magnetic force microscope (MFM) is established as a valuable tool for the analysis of magnetic structures. The standard design of MFM incorporates a silicon tip coated with a magnetic material. However, these tips are subject to several inherent problems, e.g. changing characteristics over time due to damage or magnetic hysteresis. A new theoretical electromagnetic MFM probe is introduced here. Although electromagnetic MFM has been discussed before by Zhou et al. (J. Vac. Sci. Technol. A 17 (1999) 2233), the design presented here is a different approach. Two different probe iterations and their magnetic field intensity distribution are modelled. The probe imaging capability is compared using the reciprocity principle (Wright and Hill, Appl. Phys. Lett. 68 (1996) 1726) to image the simulated force interaction between a sample and the probe fields. Thus, images of a sample's magnetic distribution are produced by the convolution of the different probe gradient field distributions and the sample magnetisation. Both perpendicular and longitudinal magnetisation patterns were simulated with the different probe iterations. This clearly showed the improvement of the second probe iteration, particularly for longitudinal patterns. The practical use of the new probe is also discussed, and future work outlined.

ITER-FEAT safety

International Nuclear Information System (INIS)

Gordon, C.W.; Bartels, H.-W.; Honda, T.; Raeder, J.; Topilski, L.; Iseli, M.; Moshonas, K.; Taylor, N.; Gulden, W.; Kolbasov, B.; Inabe, T.; Tada, E.

2001-01-01

Safety has been an integral part of the design process for ITER since the Conceptual Design Activities of the project. The safety approach adopted in the ITER-FEAT design and the complementary assessments underway, to be documented in the Generic Site Safety Report (GSSR), are expected to help demonstrate the attractiveness of fusion and thereby set a good precedent for future fusion power reactors. The assessments address ITER's radiological hazards taking into account fusion's favourable safety characteristics. The expectation that ITER will need regulatory approval has influenced the entire safety design and assessment approach. This paper summarises the ITER-FEAT safety approach and assessments underway. (author)

Seismic analysis of ITER fourth PF (Poloidal Field Coil) feeder

International Nuclear Information System (INIS)

Liu, Sumei; Chen, Wei; Song, Yuntao; Ni, Xiaojun; Wang, Zhongwei; Chen, Yonghua; Gong, Chenyu

2014-01-01

The ITER feeder systems connect the ITER magnet systems located inside the main cryostat to the cryo-plant, power-supply and control system interfaces outside the cryostat. The main purpose of the feeders is to convey the cryogenic supply and electrical power to the coils as well as house the instrumentation wiring. The PF busbar which carries 52 kA current will suffer from high Lorentz force due to the background magnetic field inspired by the coils and the self-field between every pair of busbars. Except their mechanical strength and thermal insulation performance must be achieved, the dynamic mechanism on PF structure should be assessed. This paper presents the simulation and seismic analysis on ITER 4th PF feeder including the Coil Terminal Box and S-bend Box (CTB and SBB), the Cryostat Feed-through (CFT), the In-Cryostat-Feeder (ICF), especially for the ground supports and main outer-tube firstly. This analysis aims to study seismic resistance on system design under local seismograms with floor response spectrum, the structural response vibration mode and response duration results of displacement, membrane stress, and bending stress on structure under different directions actuating signals were obtained by using the single-seismic spectrum analysis and Dead Weight analysis respectively. Based on the simulative and analytical results, the system seismic resistance and the integrity of the support structure in the 4th PF feeder have been studied and the detail design confirmed

Analytical model and design of spoke-type permanent-magnet machines accounting for saturation and nonlinearity of magnetic bridges

Energy Technology Data Exchange (ETDEWEB)

Liang, Peixin; Chai, Feng [State Key Laboratory of Robotics and System, Harbin Institute of Technology, Harbin 150001 (China); Department of Electrical Engineering, Harbin Institute of Technology, Harbin 150001 (China); Bi, Yunlong [Department of Electrical Engineering, Harbin Institute of Technology, Harbin 150001 (China); Pei, Yulong, E-mail: peiyulong1@163.com [Department of Electrical Engineering, Harbin Institute of Technology, Harbin 150001 (China); Cheng, Shukang [State Key Laboratory of Robotics and System, Harbin Institute of Technology, Harbin 150001 (China); Department of Electrical Engineering, Harbin Institute of Technology, Harbin 150001 (China)

2016-11-01

Based on subdomain model, this paper presents an analytical method for predicting the no-load magnetic field distribution, back-EMF and torque in general spoke-type motors with magnetic bridges. Taking into account the saturation and nonlinearity of magnetic material, the magnetic bridges are equivalent to fan-shaped saturation regions. For getting standard boundary conditions, a lumped parameter magnetic circuit model and iterative method are employed to calculate the permeability. The final field domain is divided into five types of simple subdomains. Based on the method of separation of variables, the analytical expression of each subdomain is derived. The analytical results of the magnetic field distribution, Back-EMF and torque are verified by finite element method, which confirms the validity of the proposed model for facilitating the motor design and optimization. - Highlights: • The no-load magnetic field of poke-type motors is firstly calculated by analytical method. • The magnetic circuit model and iterative method are employed to calculate the permeability. • The analytical expression of each subdomain is derived.. • The proposed method can effectively reduce the predesign stages duration.

Analytical model and design of spoke-type permanent-magnet machines accounting for saturation and nonlinearity of magnetic bridges

International Nuclear Information System (INIS)

Liang, Peixin; Chai, Feng; Bi, Yunlong; Pei, Yulong; Cheng, Shukang

2016-01-01

Based on subdomain model, this paper presents an analytical method for predicting the no-load magnetic field distribution, back-EMF and torque in general spoke-type motors with magnetic bridges. Taking into account the saturation and nonlinearity of magnetic material, the magnetic bridges are equivalent to fan-shaped saturation regions. For getting standard boundary conditions, a lumped parameter magnetic circuit model and iterative method are employed to calculate the permeability. The final field domain is divided into five types of simple subdomains. Based on the method of separation of variables, the analytical expression of each subdomain is derived. The analytical results of the magnetic field distribution, Back-EMF and torque are verified by finite element method, which confirms the validity of the proposed model for facilitating the motor design and optimization. - Highlights: • The no-load magnetic field of poke-type motors is firstly calculated by analytical method. • The magnetic circuit model and iterative method are employed to calculate the permeability. • The analytical expression of each subdomain is derived.. • The proposed method can effectively reduce the predesign stages duration.

Magnetic and electrical properties of ITER vacuum vessel steels

International Nuclear Information System (INIS)

Mergia, K.; Apostolopoulos, G.; Gjoka, M.; Niarchos, D.

2007-01-01

Full text of publication follows: Ferritic steel AISI 430 is a candidate material for the lTER vacuum vessel which will be used to limit the ripple in the toroidal magnetic field. The magnetic and electrical properties and their temperature dependence in the temperature range 300 - 900 K of AISI 430 ferritic stainless steels are presented. The temperature variation of the coercive field, remanence and saturation magnetization as well as electrical resistivity and the effect of annealing on these properties is discussed. (authors)

Qualification of Fin-Type Heat Exchangers for the ITER Current Leads

CERN Document Server

Ballarino, A; Bordini, B; Devred, A; Ding, K; Niu, E; Sitko, M; Taylor, T; Yang, Y; Zhou, T

2015-01-01

The ITER current leads will transfer large currents of up to 68 kA into the biggest superconducting magnets ever built. Following the development of prototypes and targeted trials of specific manufacturing processes through mock-ups, the ASIPP (Chinese Institute of Plasma Physics) is preparing for the series fabrication. A key component of the ITER HTS current leads are the resistive heat exchangers. Special R&D was conducted for these components at CERN and ASIPP in support of their designs. In particular several mock-ups were built and tested in room temperature gas to measure the dynamic pressure drop and compare to 3D CFD models.

Plasma flow to a surface using the iterative Monte Carlo method

International Nuclear Information System (INIS)

Pitcher, C.S.

1994-01-01

The iterative Monte Carlo (IMC) method is applied to a number of one-dimensional plasma flow problems, which encompass a wide range of conditions typical of those present in the boundary of magnetic fusion devices. The kinetic IMC method of solving plasma flow to a surface consists of launching and following particles within a grid of 'bins' into which weights are left according to the time a particle spends within a bin. The density and potential distributions within the plasma are iterated until the final solution is arrived at. The IMC results are compared with analytical treatments of these problems and, in general, good agreement is obtained. (author)

Intense diagnostic neutral beam development for ITER

International Nuclear Information System (INIS)

Rej, D.J.; Henins, I.; Fonck, R.J.; Kim, Y.J.

1992-01-01

For the next-generation, burning tokamak plasmas such as ITER, diagnostic neutral beams and beam spectroscopy will continue to be used to determine a variety of plasma parameters such as ion temperature, rotation, fluctuations, impurity content, current density profile, and confined alpha particle density and energy distribution. Present-day low-current, long-pulse beam technology will be unable to provide the required signal intensities because of higher beam attenuation and background bremsstrahlung radiation in these larger, higher-density plasmas. To address this problem, we are developing a short-pulse, intense diagnostic neutral beam. Protons or deuterons are accelerated using magnetic-insulated ion-diode technology, and neutralized in a transient gas cell. A prototype 25-kA, 100-kV, 1-μs accelerator is under construction at Los Alamos. Initial experiments will focus on ITER-related issues of beam energy distribution, current density, pulse length, divergence, propagation, impurity content, reproducibility, and maintenance

ITER - TVPS remote handling critical design issues

International Nuclear Information System (INIS)

1990-09-01

This report describes critical design issues concerning remote maintenance of the ITER Torus Vacuum Pumping System (TVPS). The key issues under investigation are the regeneration/isolation valve seal and seal mechanism replacement; impact of inert gas operation; impact of remote handling (RH) on the building configuration and RH equipment requirements. Seal exchange concepts are developed and their impact on the valve design identified. Concerns regarding the design and operation of RH equipment in an inert gas atmosphere are also explored. The report compares preliminary RH equipment options, pumping equipment maintenance frequency and their impact on the building design, and makes recommendations where a conflict exists between pumping equipment and the building layout. (51 figs., 11 refs.)

Role of Outgassing of ITER Vacuum Vessel In-Wall Shielding Materials in Leak Detection of ITER Vacuum Vessel

Science.gov (United States)

Maheshwari, A.; Pathak, H. A.; Mehta, B. K.; Phull, G. S.; Laad, R.; Shaikh, M. S.; George, S.; Joshi, K.; Khan, Z.

2017-04-01

ITER Vacuum Vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with In-Wall Shielding Blocks (IWS) and Water. The main purpose of IWS is to provide neutron shielding during ITER plasma operation and to reduce ripple of Toroidal Magnetic Field (TF). Although In-Wall Shield Blocks (IWS) will be submerged in water in between the walls of the ITER Vacuum Vessel (VV), Outgassing Rate (OGR) of IWS materials plays a significant role in leak detection of Vacuum Vessel of ITER. Thermal Outgassing Rate of a material critically depends on the Surface Roughness of material. During leak detection process using RGA equipped Leak detector and tracer gas Helium, there will be a spill over of mass 3 and mass 2 to mass 4 which creates a background reading. Helium background will have contribution of Hydrogen too. So it is necessary to ensure the low OGR of Hydrogen. To achieve an effective leak test it is required to obtain a background below 1 × 10-8 mbar 1 s-1 and hence the maximum Outgassing rate of IWS Materials should comply with the maximum Outgassing rate required for hydrogen i.e. 1 x 10-10 mbar 1 s-1 cm-2 at room temperature. As IWS Materials are special materials developed for ITER project, it is necessary to ensure the compliance of Outgassing rate with the requirement. There is a possibility of diffusing the gasses in material at the time of production. So, to validate the production process of materials as well as manufacturing of final product from this material, three coupons of each IWS material have been manufactured with the same technique which is being used in manufacturing of IWS blocks. Manufacturing records of these coupons have been approved by ITER-IO (International Organization). Outgassing rates of these coupons have been measured at room temperature and found in acceptable limit to obtain the required Helium Background. On the basis of these measurements, test reports have been generated and got

Nuclear systems and testing programs for ITER. Progress report for FY 1998

International Nuclear Information System (INIS)

1998-01-01

The effort during this performance period focused on a number of TBWG activities (including test module design and analysis) that were identified and agreed upon (in the presence of the ITER Director and Deputy Director) at TBWG-4. These include: (a) DEMO test module design and performance analysis under pulsed operation; (b) Test program operation plan; (c) Test port design and analysis; (d) Decay heat calculations and safety analysis; (e) Further discussion among the parties to define collaboratory on R and D for the test program as well as possible collaboration on the construction and operation of test articles; (f) Remote handling and ancillary equipment; (g) Criteria for qualifying a blanket module or submodule for actual insertion and testing in ITER; (h) Definition of test module instrumentation and verification of capability to perform in the ITER fusion environment (magnetic field, radiation, heating, etc.); and (i) Analysis to show that the results to be obtained from the test modules as designed can be extrapolated to DEMO and reactor conditions (e.g., higher wall loads and the need to demonstrate tritium self-sufficiency). The main achievements during this performance period include: (1) updating and finalizing the US DDDs for the ITER Blanket Program to form part of the ITER Final Design Report (FDR). Specific revisions were in response to the minimal lithium volume test blanket design requirements and safety impact and (2) evaluating the feasibility of the US test program, including instrumentation and the benefits of the ITER test program. Details of this assessment, including solid breeder and liquid breeder blanket test plans, are documented in UCLA-IFNT-13 (attached). In addition, dose mapping calculations were performed for the ITER Building, including equipment and layout of coolant pipes/heat exchangers. A report on ITER Building dose calculations was sent to UD ITER management and to the Garching Task Coordinator in April, 1998. The report

The physics role of ITER

International Nuclear Information System (INIS)

Rutherford, P.H.

1997-04-01

Experimental research on the International Thermonuclear Experimental Reactor (ITER) will go far beyond what is possible on present-day tokamaks to address new and challenging issues in the physics of reactor-like plasmas. First and foremost, experiments in ITER will explore the physics issues of burning plasmas--plasmas that are dominantly self-heated by alpha-particles created by the fusion reactions themselves. Such issues will include (i) new plasma-physical effects introduced by the presence within the plasma of an intense population of energetic alpha particles; (ii) the physics of magnetic confinement for a burning plasma, which will involve a complex interplay of transport, stability and an internal self-generated heat source; and (iii) the physics of very-long-pulse/steady-state burning plasmas, in which much of the plasma current is also self-generated and which will require effective control of plasma purity and plasma-wall interactions. Achieving and sustaining burning plasma regimes in a tokamak necessarily requires plasmas that are larger than those in present experiments and have higher energy content and power flow, as well as much longer pulse length. Accordingly, the experimental program on ITER will embrace the study of issues of plasma physics and plasma-materials interactions that are specific to a reactor-scale fusion experiment. Such issues will include (i) confinement physics for a tokamak in which, for the first time, the core-plasma and the edge-plasma are simultaneously in a reactor-like regime; (ii) phenomena arising during plasma transients, including so-called disruptions, in regimes of high plasma current and thermal energy; and (iii) physics of a radiative divertor designed for handling high power flow for long pulses, including novel plasma and atomic-physics effects as well as materials science of surfaces subject to intense plasma interaction. Experiments on ITER will be conducted by researchers in control rooms situated at major

Conceptual design of the ITER fast-ion loss detector

International Nuclear Information System (INIS)

Garcia-Munoz, M.; Ayllon-Guerola, J.; Galdon, J.; Garcia Lopez, J.; Gonzalez-Martin, J.; Jimenez-Ramos, M. C.; Rodriguez-Ramos, M.; Rivero-Rodriguez, J. F.; Sanchis-Sanchez, L.; Kocan, M.; Bertalot, L.; Bonnet, Y.; Casal, N.; Giacomin, T.; Pinches, S. D.; Reichle, R.; Vayakis, G.; Veshchev, E.; Vorpahl, Ch.; Walsh, M.

2016-01-01

A conceptual design of a reciprocating fast-ion loss detector for ITER has been developed and is presented here. Fast-ion orbit simulations in a 3D magnetic equilibrium and up-to-date first wall have been carried out to revise the measurement requirements for the lost alpha monitor in ITER. In agreement with recent observations, the simulations presented here suggest that a pitch-angle resolution of ∼5° might be necessary to identify the loss mechanisms. Synthetic measurements including realistic lost alpha-particle as well as neutron and gamma fluxes predict scintillator signal-to-noise levels measurable with standard light acquisition systems with the detector aperture at ∼11 cm outside of the diagnostic first wall. At measurement position, heat load on detector head is comparable to that in present devices.

Conceptual design of the ITER fast-ion loss detector

Energy Technology Data Exchange (ETDEWEB)

Garcia-Munoz, M., E-mail: mgm@us.es; Ayllon-Guerola, J.; Galdon, J.; Garcia Lopez, J.; Gonzalez-Martin, J.; Jimenez-Ramos, M. C.; Rodriguez-Ramos, M.; Rivero-Rodriguez, J. F.; Sanchis-Sanchez, L. [Department of Atomic, Molecular and Nuclear Physics, University of Seville, 41012 Seville (Spain); CNA (Universidad de Sevilla-CSIC-J. Andalucía), Seville (Spain); Kocan, M.; Bertalot, L.; Bonnet, Y.; Casal, N.; Giacomin, T.; Pinches, S. D.; Reichle, R.; Vayakis, G.; Veshchev, E.; Vorpahl, Ch.; Walsh, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); and others

2016-11-15

A conceptual design of a reciprocating fast-ion loss detector for ITER has been developed and is presented here. Fast-ion orbit simulations in a 3D magnetic equilibrium and up-to-date first wall have been carried out to revise the measurement requirements for the lost alpha monitor in ITER. In agreement with recent observations, the simulations presented here suggest that a pitch-angle resolution of ∼5° might be necessary to identify the loss mechanisms. Synthetic measurements including realistic lost alpha-particle as well as neutron and gamma fluxes predict scintillator signal-to-noise levels measurable with standard light acquisition systems with the detector aperture at ∼11 cm outside of the diagnostic first wall. At measurement position, heat load on detector head is comparable to that in present devices.

ITER council proceedings: 1998

International Nuclear Information System (INIS)

1999-01-01

This volume contains documents of the 13th and the 14th ITER council meeting as well as of the 1st extraordinary ITER council meeting. Documents of the ITER meetings held in Vienna and Yokohama during 1998 are also included. The contents include an outline of the ITER objectives, the ITER parameters and design overview as well as operating scenarios and plasma performance. Furthermore, design features, safety and environmental characteristics are given

The development of beryllium plasma spray technology for the International Thermonuclear Experimental Reactor (ITER)

International Nuclear Information System (INIS)

Castro, R.G.; Elliott, K.E.; Hollis, K.J.; Watson, R.D.

1999-01-01

Over the past five years, four international parties, which include the European Communities, Japan, the Russian Federation and the United States, have been collaborating on the design and development of the International Thermonuclear Experimental Reactor (ITER), the next generation magnetic fusion energy device. During the ITER Engineering Design Activity (EDA), beryllium plasma spray technology was investigated by Los Alamos National Laboratory as a method for fabricating and repairing and the beryllium first wall surface of the ITER tokamak. Significant progress has been made in developing beryllium plasma spraying technology for this application. Information will be presented on the research performed to improve the thermal properties of plasma sprayed beryllium coatings and a method that was developed for cleaning and preparing the surface of beryllium prior to depositing plasma sprayed beryllium coatings. Results of high heat flux testing of the beryllium coatings using electron beam simulated ITER conditions will also be presented

The administration of psilocybin to healthy, hallucinogen-experienced volunteers in a mock-functional magnetic resonance imaging environment: a preliminary investigation of tolerability.

Science.gov (United States)

Carhart-Harris, Robin L; Williams, Tim M; Sessa, Ben; Tyacke, Robin J; Rich, Ann S; Feilding, Amanda; Nutt, David J

2011-11-01

This study sought to assess the tolerability of intravenously administered psilocybin in healthy, hallucinogen-experienced volunteers in a mock-magnetic resonance imaging environment as a preliminary stage to a controlled investigation using functional magnetic resonance imaging to explore the effects of psilocybin on cerebral blood flow and activity. The present pilot study demonstrated that up to 2 mg of psilocybin delivered as a slow intravenous injection produces short-lived but typical drug effects that are psychologically and physiologically well tolerated. With appropriate care, this study supports the viability of functional magnetic resonance imaging work with psilocybin.

Iter

Science.gov (United States)

Iotti, Robert

2015-04-01

ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

Design meeting on reduced technical objectives/reduced cost ITER options

International Nuclear Information System (INIS)

Spears, W.

1999-01-01

At this meeting, which took place at Garching, Germany in January 1999, means of reducing the overall cost for ITER to 50% where discussed. It was felt that a smaller plasma of high elongation and high triangularity was a step in the right direction. Further steps would include cheaper magnetic field coils, cheaper in-vessel components and also costly buildings

Key enabling design features of the ITER HNB Duct Liner

Energy Technology Data Exchange (ETDEWEB)

Chuilon, Ben, E-mail: ben.chuilon@ccfe.ac.uk; Mistry, Sanjay; Andrews, Rodney; Verhoeven, Roel; Xue, Yongkuan

2015-10-15

Highlights: • Key engineering design details of the ITER HND Duct Liner are presented. • A standardised CuCrZr water cooled panel that can be remotely handled is detailed. • Bolts are protected from beam power by means of a tungsten cap to radiate heat away. • Water connections placed coaxially are protected from beam power by a tungsten ring. • Explosion-bonded CuCrZr-316L panels result in a tenfold disruption torque reduction. - Abstract: The Duct Liner (DL) for the ITER Heating Neutral Beam (HNB) is a key component in the beam transport system. Duct Liners installed into equatorial ports 4 and 5 of the Vacuum Vessel (VV) will protect the port extension from power deposition due to re-ionisation and direct interception of the HNB. Furthermore, the DL contributes towards the shielding of the VV and superconducting coils from plasma photons and neutrons. The DL incorporates a 316L(N)-IG, deep-drilled and water cooled Neutron Shield (NS) whose internal walls are lined with actively cooled CuCrZr Duct Liner Modules (DLMs). These Remote Handling Class 2 and 3 panels provide protection from neutral beam power. This paper provides an overview of the preliminary design for the ITER HNB DL and focusses on critical features that ensure compatibility with: high heat flux requirements, remote maintenance procedures, and transient magnetic fields arising from major plasma disruptions. The power deposited on a single DLM can reach 300 kW with a peak power density of 2.4 MW/m{sup 2}. Feeding coolant to the DLMs is accomplished via welded connections to the internal coolant network of the NS. These are placed coaxially to allow for thermal expansion of the DLMs without the use of deformable connections. Critically, the remote maintenance of individual DLMs necessitates access to water connections and bolts from the beam facing surface, thus subjecting them to high heat flux loads. This design challenge will become more prevalent as fusion devices become more powerful

Co teď můžeme vidět na staveništi tokamaku ITER?

Czech Academy of Sciences Publication Activity Database

Řípa, Milan

Listopad (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER site * Assembly Hall,Radio Frequency Building * Tritium Building * Tokamak Building * Diagnostic Building * Magnet Power Conversion Building * 400 kV Switchyard * Cryoplant * Coils Winding Facility Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) https://www.3pol.cz/cz/rubriky/jaderna-fyzika-a-energetika/2084-co-ted-muzeme-videt-na-stavenisti-tokamaku-iter

Measurement of tokamak error fields using plasma response and its applicability to ITER

International Nuclear Information System (INIS)

Strait, E.J.; Buttery, R.J.; Chu, M.S.; Garofalo, A.M.; La Haye, R.J.; Schaffer, M.J.; Casper, T.A.; Gribov, Y.; Hanson, J.M.; Reimerdes, H.; Volpe, F.A.

2014-01-01

The nonlinear response of a low-beta tokamak plasma to non-axisymmetric fields offers an alternative to direct measurement of the non-axisymmetric part of the vacuum magnetic fields, often termed ‘error fields’. Possible approaches are discussed for determination of error fields and the required current in non-axisymmetric correction coils, with an emphasis on two relatively new methods: measurement of the torque balance on a saturated magnetic island, and measurement of the braking of plasma rotation in the absence of an island. The former is well suited to ohmically heated discharges, while the latter is more appropriate for discharges with a modest amount of neutral beam heating to drive rotation. Both can potentially provide continuous measurements during a discharge, subject to the limitation of a minimum averaging time. The applicability of these methods to ITER is discussed, and an estimate is made of their uncertainties in light of the specifications of ITER's diagnostic systems. The use of plasma response-based techniques in normal ITER operational scenarios may allow identification of the error field contributions by individual central solenoid coils, but identification of the individual contributions by the outer poloidal field coils or other sources is less likely to be feasible. (paper)

Iterative CT reconstruction with correction for known rigid motion

Energy Technology Data Exchange (ETDEWEB)

Nuyts, Johan [Katholieke Univ. Leuven (Belgium). Dept. of Nuclear Medicine; Kim, Jung-Ha; Fulton, Roger [Sydney Univ., NSW (Australia). School of Physics; Westmead Hospital, Sydney (Australia). Medical Physics

2011-07-01

In PET/CT brain imaging, correction for motion may be needed, in particular for children and psychiatric patients. Motion is more likely to occur in the lengthy PET measurement, but also during the short CT acquisition patient motion is possible. Rigid motion of the head can be measured independently from the PET/CT system with optical devices. In this paper, we propose a method and some preliminary simulation results for iterative CT reconstruction with correction for known rigid motion. We implemented an iterative algorithm for fully 3D reconstruction from helical CT scans. The motion of the head is incorporated in the system matrix as a view-dependent motion of the CT-system. The first simulation results indicate that some motion patterns may produce loss of essential data. This loss precludes exact reconstruction and results in artifacts in the reconstruction, even when motion is taken into account. However, by reducing the pitch during acquisition, the same motion pattern no longer caused artifacts in the motion corrected image. (orig.)

ITER test programme

International Nuclear Information System (INIS)

Abdou, M.; Baker, C.; Casini, G.

1991-01-01

ITER has been designed to operate in two phases. The first phase which lasts for 6 years, is devoted to machine checkout and physics testing. The second phase lasts for 8 years and is devoted primarily to technology testing. This report describes the technology test program development for ITER, the ancillary equipment outside the torus necessary to support the test modules, the international collaboration aspects of conducting the test program on ITER, the requirements on the machine major parameters and the R and D program required to develop the test modules for testing in ITER. 15 refs, figs and tabs

ITER-FEAT outline design report

International Nuclear Information System (INIS)

2001-01-01

In July 1998 the ITER Parties were unable, for financial reasons, to proceed with construction of the ITER design proposed at that time, to meet the detailed technical objectives and target cost set in 1992. It was therefore decided to investigate options for the design of ITER with reduced technical objectives and with possibly decreased technical margins, whose target construction cost was one half that of the 1998 ITER design, while maintaining the overall programmatic objective. To identify designs that might meet the revised objectives, task forces involving the JCT and Home Teams met during 1998 and 1999 to analyse and compare a range of options for the design of such a device. This led at the end of 1999 to a single configuration for the ITER design with parameters considered to be the most credible consistent with technical limitations and the financial target, yet meeting fully the objectives with appropriate margins. This new design of ITER, called ''ITER-FEAT'', was submitted to the ITER Director to the ITER Parties as the ''ITER-FEAT Outline Design Report'' (ODR) in January 2000, at their meeting in Tokyo. The Parties subsequently conducted their domestic assessments of this report and fed the resulting comments back into the progressing design. The progress on the developing design was reported to the ITER Technical Advisory Committee (TAC) in June 2000 in the report ''Progress in Resolving Open Design Issues from the ODR'' alongside a report on Progress in Technology R and D for ITER. In addition, the progress in the ITER-FEAT Design and Validating R and D was reported to the ITER Parties. The ITER-FEAT design was subsequently approved by the governing body of ITER in Moscow in June 2000 as the basis for the preparation of the Final Design Report, recognising it as a single mature design for ITER consistent with its revised objectives. This volume contains the documents pertinent to the process described above. More detailed technical information

Progress and present status of ITER cryoline system

International Nuclear Information System (INIS)

Badgujar, S.; Bonneton, M.; Chalifour, M.; Forgeas, A.; Serio, L.; Sarkar, B.; Shah, N.

2014-01-01

The cryoline system at ITER forms a very complex network localized inside the Tokamak building, on a dedicated plant bridge and in cryoplant areas. The cooling power produced in the cryoplant is distributed via these lines with a total length of about 3.7 km and interconnecting all the cold boxes of the cryogenic system as well as the cold boxes of various clients (magnets, cryopumps and thermal shield). Distinct layouts and polygonal geometry, nuclear safety and confinement requirements, difficult installation and in-service inspection/repair demand very high reliability and availability for the cryolines. The finalization of the building-embedded plates for supporting the lines, before the detailed design, has made this project technologically more challenging. The conceptual design phase has been completed and procurement arrangements have been signed with India, responsible for providing the system of cryolines and warm lines to ITER, as in kind contribution. The prototype test for the design and performance validation has been planned on a representative cryoline section. After describing the basic features and general layout of the ITER cryolines, the paper presents key design requirements, conceptual design approach, progress and status of the cryolines project as well as challenges to build such a complex cryoline system

Magnetics and superconductivity section annual progress report for period ending December 31, 1976

International Nuclear Information System (INIS)

Lubell, M.S.; Dresner, L.

1977-06-01

The Magnetics and Superconductivity Section has the responsibility for developing superconducting magnet systems for tokamak fusion machines. This is being accomplished by carrying out those research and development needs which will provide the physics understanding and engineering data necessary to design, fabricate, and test large toroidal field (TF) and poloidal field (PF) coils. This information, in addition, supports the Large Coil Program (LCP). A number of design projects have been performed, some in support of other programs and some of a continuing nature. These efforts support the goals and requirements for both the TF and PF magnet systems. Examples are the magnet designs for the EPR, Demo, EBTR, EBT-II, and preliminary scoping for the INS project. The principal effort was expended on the iteration of the EPR Reference Design. Three features of the original reference design--the honeycomb coil structure, the oval coil shape, and the forced-flow cooling of the conductor by supercritical helium--remain as key features of the TF coils. Considerable progress has been made in the theoretical understanding of forced-flow-cooled conductors, and optimized designs with maximum stability margin can be designed to meet specific applications. Experiments which will test the theory are in progress

ITER council proceedings: 1995

International Nuclear Information System (INIS)

1996-01-01

Records of the 8. ITER Council Meeting (IC-8), held on 26-27 July 1995, in San Diego, USA, and the 9. ITER Council Meeting (IC-9) held on 12-13 December 1995, in Garching, Germany, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the ITER Interim Design Report Package and Relevant Documents. Figs, tabs

Acquisition procedures, processing methodologies and preliminary results of magnetic and ROV data collected during the TOMO-ETNA experiment

Directory of Open Access Journals (Sweden)

Danilo Cavallaro

2016-09-01

Full Text Available The TOMO-ETNA experiment was devised for the investigation of the continental and oceanic crust beneath Mt. Etna volcano and northeastern Sicily up to the Aeolian Islands, through an active source study. In this experiment, a large amount of geophysical data was collected both inland and in the Ionian and Tyrrhenian Seas for identifying the major geological and structural features offshore Mt. Etna and NE Sicily. One of the oceanographic cruises organized within the TOMO-ETNA experiment was carried out on the hydrographic vessel “Galatea” by Italian Navy. During the cruise a detailed magnetic survey and a set of ROV (remotely operated vehicle dives were performed offshore Mt. Etna. The magnetic survey allowed the compilation of a preliminary magnetic map revealing a clear direct relationship between volcanic structures and high frequency magnetic anomalies. Significant positive magnetic anomalies were identified offshore the Timpa area and along the easternmost portion of the Riposto Ridge and correlated to a primitive volcanic edifice and to shallow volcanic bodies, respectively. On the whole, the magnetic anomaly map highlights a clear SW-NE decreasing trend, where high amplitude positive magnetic anomaly pattern of the SW sector passes, northeastwardly, to a main negative one. ROV dives permitted to directly explore the shallowest sectors of the Riposto Ridge and to collect several videos and seafloor samples, allowing us to identify some locally developed volcanic manifestations.

Tissue characterization using magnetic resonance elastography: preliminary results

International Nuclear Information System (INIS)

Kruse, S.A.; Smith, J.A.; Lawrence, A.J.; Dresner, M.A.; Manduca, A.; Greenleaf, J.F.; Ehman, R.L.

2000-01-01

The well-documented effectiveness of palpation as a diagnostic technique for detecting cancer and other diseases has provided motivation for developing imaging techniques for non-invasively evaluating the mechanical properties of tissue. A recently described approach for elasticity imaging, using propagating acoustic shear waves and phase-contrast MRI, has been called magnetic resonance elastography (MRE). The purpose of this work was to conduct preliminary studies to define methods for using MRE as a tool for addressing the paucity of quantitative tissue mechanical property data in the literature. Fresh animal liver and kidney tissue specimens were evaluated with MRE at multiple shear wave frequencies. The influence of specimen temperature and orientation on measurements of stiffness was studied in skeletal muscle. The results demonstrated that all of the materials tested (liver, kidney, muscle and tissue-simulating gel) exhibit systematic dependence of shear stiffness on shear rate. These data are consistent with a viscoelastic model of tissue mechanical properties, allowing calculation of two independent tissue properties from multiple-frequency MRE data: shear modulus and shear viscosity. The shear stiffness of tissue can be substantially affected by specimen temperature. The results also demonstrated evidence of shear anisotropy in skeletal muscle but not liver tissue. The measured shear stiffness in skeletal muscle was found to depend on both the direction of propagation and polarization of the shear waves. (author)

submitter Starting Manufacture of the ITER Central Solenoid

CERN Document Server

Libeyre, P; Dolgetta, N; Gaxiola, E; Jong, C; Lyraud, C; Mitchell, N; Journeaux, J Y; Vollmann, T; Evans, D; Sgobba, S; Langeslag, S; Reiersen, W; Martovetsky, N; Everitt, D; Hatfield, D; Rosenblad, P; Litherland, S; Freudenberg, K; Myatt, L; Smith, J; Brazelton, C; Abbott, R; Daubert, J; Rackers, K; Nentwich, T

2016-01-01

The central solenoid (CS) is a key component of the ITER magnet system to provide the magnetic flux swing required to drive induced plasma current up to 15 MA. The manufacture of its different subcomponents has now started, following completion of the design analyses and achievement of the qualification of the manufacturing procedures. A comprehensive set of analyses has been produced to demonstrate that the CS final design meets all requirements. This includes in particular structural analyses carried out with different finite-element models and addressing normal and fault conditions. Following the Final Design Review, held in November 2013, and the subsequent design modifications, the analyses were updated for consistency with the final design details and provide evidence that the Magnet Structural Design Criteria are fully met. Before starting any manufacturing activity of a CS component, a corresponding dedicated qualification program has been carried out. This includes manufacture of mockups using the re...

A Brownian dynamics study on ferrofluid colloidal dispersions using an iterative constraint method to satisfy Maxwell’s equations

Energy Technology Data Exchange (ETDEWEB)

Dubina, Sean Hyun, E-mail: sdubin2@uic.edu; Wedgewood, Lewis Edward, E-mail: wedge@uic.edu [Department of Chemical Engineering, University of Illinois at Chicago, 810 S. Clinton St. (MC 110), Chicago, Illinois 60607-4408 (United States)

2016-07-15

Ferrofluids are often favored for their ability to be remotely positioned via external magnetic fields. The behavior of particles in ferromagnetic clusters under uniformly applied magnetic fields has been computationally simulated using the Brownian dynamics, Stokesian dynamics, and Monte Carlo methods. However, few methods have been established that effectively handle the basic principles of magnetic materials, namely, Maxwell’s equations. An iterative constraint method was developed to satisfy Maxwell’s equations when a uniform magnetic field is imposed on ferrofluids in a heterogeneous Brownian dynamics simulation that examines the impact of ferromagnetic clusters in a mesoscale particle collection. This was accomplished by allowing a particulate system in a simple shear flow to advance by a time step under a uniformly applied magnetic field, then adjusting the ferroparticles via an iterative constraint method applied over sub-volume length scales until Maxwell’s equations were satisfied. The resultant ferrofluid model with constraints demonstrates that the magnetoviscosity contribution is not as substantial when compared to homogeneous simulations that assume the material’s magnetism is a direct response to the external magnetic field. This was detected across varying intensities of particle-particle interaction, Brownian motion, and shear flow. Ferroparticle aggregation was still extensively present but less so than typically observed.

A Brownian dynamics study on ferrofluid colloidal dispersions using an iterative constraint method to satisfy Maxwell’s equations

International Nuclear Information System (INIS)

Dubina, Sean Hyun; Wedgewood, Lewis Edward

2016-01-01

Ferrofluids are often favored for their ability to be remotely positioned via external magnetic fields. The behavior of particles in ferromagnetic clusters under uniformly applied magnetic fields has been computationally simulated using the Brownian dynamics, Stokesian dynamics, and Monte Carlo methods. However, few methods have been established that effectively handle the basic principles of magnetic materials, namely, Maxwell’s equations. An iterative constraint method was developed to satisfy Maxwell’s equations when a uniform magnetic field is imposed on ferrofluids in a heterogeneous Brownian dynamics simulation that examines the impact of ferromagnetic clusters in a mesoscale particle collection. This was accomplished by allowing a particulate system in a simple shear flow to advance by a time step under a uniformly applied magnetic field, then adjusting the ferroparticles via an iterative constraint method applied over sub-volume length scales until Maxwell’s equations were satisfied. The resultant ferrofluid model with constraints demonstrates that the magnetoviscosity contribution is not as substantial when compared to homogeneous simulations that assume the material’s magnetism is a direct response to the external magnetic field. This was detected across varying intensities of particle-particle interaction, Brownian motion, and shear flow. Ferroparticle aggregation was still extensively present but less so than typically observed.

Magnetic filter field for ELISE––Concepts and design

International Nuclear Information System (INIS)

Fröschle, M.; Fantz, U.; Franzen, P.; Kraus, W.; Nocentini, R.; Schiesko, L.; Wünderlich, D.

2013-01-01

Highlights: ► ELISE is an important intermediate step toward the full size ITER injector ion source ► It is one of the first ion sources equipped with a magnetic filter field formed by a PG current. ► The magnetic filter field is responsible for the performance of the source ► It controls the currents of extracted negative ions and co-extracted electrons ► The ELISE magnetic filter field meets all actual scientific findings ► It has a vast variability for future investigations and optimizations. -- Abstract: Negative ion neutral beam injection heating systems as planned for ITER need efficient precautions in the plasma source to minimize the co-extraction of electrons and destruction of negative ions. One solution is to apply a magnetic filter field of several mT, which reduces the electron temperature and the amount of electrons in the extraction region in front of the plasma grid. For the small IPP prototype sources it has been found, that both, the absolute value of the magnetic flux density in the extraction region as well as its integral along the distance from plasma driver to plasma grid has an important influence on the performance of the source. In the ITER ion sources, a strong current of several kA driven through the plasma grid is used to create the transversal magnetic field. The test bed ELISE (Extraction from a Large Ion Source Experiment) at IPP Garching houses the first negative ion source with the full width of the ITER source, with a similar aperture arrangement of the extraction system and with a magnetic filter field formed by a plasma grid current. One issue of the research at this test facility will be to explore and optimize the magnetic filter field. The paper summarizes experiences and results of previous filter field test campaigns and presents the magnetic filter field design for ELISE

ITER CTA newsletter. No. 3

International Nuclear Information System (INIS)

2001-11-01

This ITER CTA newsletter comprises reports of Dr. P. Barnard, Iter Canada Chairman and CEO, about the progress of the first formal ITER negotiations and about the demonstration of details of Canada's bid on ITER workshops, and Dr. V. Vlasenkov, Project Board Secretary, about the meeting of the ITER CTA project board

Mechanical design and construction qualification program on ITER correction coils structures

Energy Technology Data Exchange (ETDEWEB)

Foussat, A., E-mail: arnaud.foussat@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Weiyue, Wu; Jing, Wei; Shuangsong, Du [Academy of Science Institute of Plasma Physics, PO 1126, Hefei, Anhui 230031 (China); Sgobba, S. [European Center for Nuclear Research, CH-1211 Geneva 23 (Switzerland); Hongwei, Li [China International Nuclear Fusion Energy Program Execution Center, Ministry of Science and Technology, 15B Fuxing Rd., Beijing 100862 (China); Libeyre, Paul; Jong, Cornelis; Klofac, Kamil; Mitchell, Neil [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2014-04-01

The ITER Magnet system consists of 4 main coils sub-systems, i.e. 18 toroidal field coils (TFC), a central solenoid (CS), 6 poloidal field coils (PF) and 3 sets of correction coils (CC). The ITER fusion project has selected the stainless steel 316LN as main material for the magnet structure. The CC contribute to reducing the range of magnetic error fields created by imperfections in the location and geometry of the other coils used to confine, heat, and shape the plasma. During plasma operation, a large number of loading condition scenarios have been considered and structural analysis performed on key items like Cable-In-Conduit Conductor and the coil case. The results obtained are used for both static and fatigue structural assessment defining the present baseline design. For the construction of the structural cases, welding techniques such as GTAW (Gas Tungsten Arc Welding) and techniques resulting in low distortion and shrinkage like EBW (Electron Beam Welding) or Laser Beam Welding (LBW) with filler metal wire have been selected. Those methods are considered for future qualifications to guarantee proper weld parameters and specified weld properties. In order to determine the strength and fracture toughness of 316LN stainless steel welds with respect to design criteria, some mechanical tests have been carried out at 7 K (or 77 K), and room temperature.

ITER council proceedings: 1997

International Nuclear Information System (INIS)

1997-01-01

This volume of the ITER EDA Documentation Series presents records of the 12th ITER Council Meeting, IC-12, which took place on 23-24 July, 1997 in Tampere, Finland. The Council received from the Parties (EU, Japan, Russia, US) positive responses on the Detailed Design Report. The Parties stated their willingness to contribute to fulfil their obligations in contributing to the ITER EDA. The summary discussions among the Parties led to the consensus that in July 1998 the ITER activities should proceed for additional three years with a general intent to enable an efficient start of possible, future ITER construction

Progress in XRCS-Survey plant instrumentation and control design for ITER

International Nuclear Information System (INIS)

Varshney, Sanjeev; Jha, Shivakant; Simrock, Stefan; Barnsley, Robin; Martin, Vincent; Mishra, Sapna; Patil, Prabhakant; Patel, Shreyas; Kumar, Vinay

2016-01-01

Highlights: • An identification of the major process functions system compliant to Plant Control Design Handbook (PCDH) has been made for XRCS-Survey plant I&C. • I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using Enterprise architect (EA). • I&C architecture, interface with ITER networks and Plants, configuration of cubicles are discussed towards nine design review deliverables. - Abstract: A real time, plasma impurity survey system based on X-ray Crystal Spectroscopy (XRCS) has been designed for ITER and will be made available in the set of first plasma diagnostics for measuring impurity ion concentrations and their in-flux. For the purpose of developing a component level design of XRCS-Survey plant I&C system that is compliant to the rules and guidelines defined in the Plant Control Design Handbook (PCDH), firstly an identification of the major process functions has been made. The preliminary plant I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using a system engineering tool, Enterprise Architect (EA). Conceptual I&C architecture, interface with the ITER networks and other Plants have been discussed along with the basic configuration of I&C cubicles aiming towards nine I&C deliverables for the design review.

Progress in XRCS-Survey plant instrumentation and control design for ITER

Energy Technology Data Exchange (ETDEWEB)

Varshney, Sanjeev, E-mail: sanjeev.varshney@iter-india.org [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Jha, Shivakant [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Simrock, Stefan; Barnsley, Robin; Martin, Vincent [ITER-Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul-Lez-Durance, Cedex (France); Mishra, Sapna [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Patil, Prabhakant [ITER-Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul-Lez-Durance, Cedex (France); Patel, Shreyas; Kumar, Vinay [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India)

2016-11-15

Highlights: • An identification of the major process functions system compliant to Plant Control Design Handbook (PCDH) has been made for XRCS-Survey plant I&C. • I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using Enterprise architect (EA). • I&C architecture, interface with ITER networks and Plants, configuration of cubicles are discussed towards nine design review deliverables. - Abstract: A real time, plasma impurity survey system based on X-ray Crystal Spectroscopy (XRCS) has been designed for ITER and will be made available in the set of first plasma diagnostics for measuring impurity ion concentrations and their in-flux. For the purpose of developing a component level design of XRCS-Survey plant I&C system that is compliant to the rules and guidelines defined in the Plant Control Design Handbook (PCDH), firstly an identification of the major process functions has been made. The preliminary plant I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using a system engineering tool, Enterprise Architect (EA). Conceptual I&C architecture, interface with the ITER networks and other Plants have been discussed along with the basic configuration of I&C cubicles aiming towards nine I&C deliverables for the design review.

Magnet Design Considerations for Fusion Nuclear Science Facility

Energy Technology Data Exchange (ETDEWEB)

Zhai, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kessel, C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States) Fusion Technology Institute; Titus, P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

2016-06-01

The Fusion Nuclear Science Facility (FNSF) is a nuclear confinement facility that provides a fusion environment with components of the reactor integrated together to bridge the technical gaps of burning plasma and nuclear science between the International Thermonuclear Experimental Reactor (ITER) and the demonstration power plant (DEMO). Compared with ITER, the FNSF is smaller in size but generates much higher magnetic field, i.e., 30 times higher neutron fluence with three orders of magnitude longer plasma operation at higher operating temperatures for structures surrounding the plasma. Input parameters to the magnet design from system code analysis include magnetic field of 7.5 T at the plasma center with a plasma major radius of 4.8 m and a minor radius of 1.2 m and a peak field of 15.5 T on the toroidal field (TF) coils for the FNSF. Both low-temperature superconductors (LTS) and high-temperature superconductors (HTS) are considered for the FNSF magnet design based on the state-of-the-art fusion magnet technology. The higher magnetic field can be achieved by using the high-performance ternary restacked-rod process Nb3Sn strands for TF magnets. The circular cable-in-conduit conductor (CICC) design similar to ITER magnets and a high-aspect-ratio rectangular CICC design are evaluated for FNSF magnets, but low-activation-jacket materials may need to be selected. The conductor design concept and TF coil winding pack composition and dimension based on the horizontal maintenance schemes are discussed. Neutron radiation limits for the LTS and HTS superconductors and electrical insulation materials are also reviewed based on the available materials previously tested. The material radiation limits for FNSF magnets are defined as part of the conceptual design studies for FNSF magnets.

Iterative and iterative-noniterative integral solutions in 3-loop massive QCD calculations

International Nuclear Information System (INIS)

Ablinger, J.; Radu, C.S.; Schneider, C.; Behring, A.; Imamoglu, E.; Van Hoeij, M.; Von Manteuffel, A.; Raab, C.G.

2017-11-01

Various of the single scale quantities in massless and massive QCD up to 3-loop order can be expressed by iterative integrals over certain classes of alphabets, from the harmonic polylogarithms to root-valued alphabets. Examples are the anomalous dimensions to 3-loop order, the massless Wilson coefficients and also different massive operator matrix elements. Starting at 3-loop order, however, also other letters appear in the case of massive operator matrix elements, the so called iterative non-iterative integrals, which are related to solutions based on complete elliptic integrals or any other special function with an integral representation that is definite but not a Volterra-type integral. After outlining the formalism leading to iterative non-iterative integrals,we present examples for both of these cases with the 3-loop anomalous dimension γ (2) qg and the structure of the principle solution in the iterative non-interative case of the 3-loop QCD corrections to the ρ-parameter.

Iterative and iterative-noniterative integral solutions in 3-loop massive QCD calculations

Energy Technology Data Exchange (ETDEWEB)

Ablinger, J.; Radu, C.S.; Schneider, C. [Johannes Kepler Univ., Linz (Austria). Research Inst. for Symbolic Computation (RISC); Behring, A. [RWTH Aachen Univ. (Germany). Inst. fuer Theoretische Teilchenphysik und Kosmologie; Bluemlein, J.; Freitas, A. de [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Imamoglu, E.; Van Hoeij, M. [Florida State Univ., Tallahassee, FL (United States). Dept. of Mathematics; Von Manteuffel, A. [Michigan State Univ., East Lansing, MI (United States). Dept. of Physics and Astronomy; Raab, C.G. [Johannes Kepler Univ., Linz (Austria). Inst. for Algebra

2017-11-15

Various of the single scale quantities in massless and massive QCD up to 3-loop order can be expressed by iterative integrals over certain classes of alphabets, from the harmonic polylogarithms to root-valued alphabets. Examples are the anomalous dimensions to 3-loop order, the massless Wilson coefficients and also different massive operator matrix elements. Starting at 3-loop order, however, also other letters appear in the case of massive operator matrix elements, the so called iterative non-iterative integrals, which are related to solutions based on complete elliptic integrals or any other special function with an integral representation that is definite but not a Volterra-type integral. After outlining the formalism leading to iterative non-iterative integrals,we present examples for both of these cases with the 3-loop anomalous dimension γ{sup (2)}{sub qg} and the structure of the principle solution in the iterative non-interative case of the 3-loop QCD corrections to the ρ-parameter.

Evaluation of ITER MSE Viewing Optics

International Nuclear Information System (INIS)

Allen, S; Lerner, S; Morris, K; Jayakumar, J; Holcomb, C; Makowski, M; Latkowski, J; Chipman, R

2007-01-01

The Motional Stark Effect (MSE) diagnostic on ITER determines the local plasma current density by measuring the polarization angle of light resulting from the interaction of a high energy neutral heating beam and the tokamak plasma. This light signal has to be transmitted from the edge and core of the plasma to a polarization analyzer located in the port plug. The optical system should either preserve the polarization information, or it should be possible to reliably calibrate any changes induced by the optics. This LLNL Work for Others project for the US ITER Project Office (USIPO) is focused on the design of the viewing optics for both the edge and core MSE systems. Several design constraints were considered, including: image quality, lack of polarization aberrations, ease of construction and cost of mirrors, neutron shielding, and geometric layout in the equatorial port plugs. The edge MSE optics are located in ITER equatorial port 3 and view Heating Beam 5, and the core system is located in equatorial port 1 viewing heating beam 4. The current work is an extension of previous preliminary design work completed by the ITER central team (ITER resources were not available to complete a detailed optimization of this system, and then the MSE was assigned to the US). The optimization of the optical systems at this level was done with the ZEMAX optical ray tracing code. The final LLNL designs decreased the ''blur'' in the optical system by nearly an order of magnitude, and the polarization blur was reduced by a factor of 3. The mirror sizes were reduced with an estimated cost savings of a factor of 3. The throughput of the system was greater than or equal to the previous ITER design. It was found that optical ray tracing was necessary to accurately measure the throughput. Metal mirrors, while they can introduce polarization aberrations, were used close to the plasma because of the anticipated high heat, particle, and neutron loads. These mirrors formed an intermediate

Two-dimensional over-all neutronics analysis of the ITER device

Science.gov (United States)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

Two-dimensional over-all neutronics analysis of the ITER device

International Nuclear Information System (INIS)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi.

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR) and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li 2 O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No.5 and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design. (author)

Issues in developing parallel iterative algorithms for solving partial differential equations on a (transputer-based) distributed parallel computing system

International Nuclear Information System (INIS)

Rajagopalan, S.; Jethra, A.; Khare, A.N.; Ghodgaonkar, M.D.; Srivenkateshan, R.; Menon, S.V.G.

1990-01-01

Issues relating to implementing iterative procedures, for numerical solution of elliptic partial differential equations, on a distributed parallel computing system are discussed. Preliminary investigations show that a speed-up of about 3.85 is achievable on a four transputer pipeline network. (author). 2 figs., 3 a ppendixes., 7 refs

Physics research needs for ITER

International Nuclear Information System (INIS)

Sauthoff, N.R.

1995-01-01

Design of ITER entails the application of physics design tools that have been validated against the world-wide data base of fusion research. In many cases, these tools do not yet exist and must be developed as part of the ITER physics program. ITER's considerable increases in power and size demand significant extrapolations from the current data base; in several cases, new physical effects are projected to dominate the behavior of the ITER plasma. This paper focuses on those design tools and data that have been identified by the ITER team and are not yet available; these needs serve as the basis for the ITER Physics Research Needs, which have been developed jointly by the ITER Physics Expert Groups and the ITER design team. Development of the tools and the supporting data base is an on-going activity that constitutes a significant opportunity for contributions to the ITER program by fusion research programs world-wide

Impact of risk analysis on ITER nuclear buildings layout

International Nuclear Information System (INIS)

Sabathe, Laurent; Lignini, Franck; Rodriguez-Rodrigo, Lina; Uzan-Elbez, Joelle; Girard, Jean-Philippe

2007-01-01

The internal layout of ITER nuclear buildings (Tokamak, Tritium, Hot cell and Radwaste buildings) is the result of an iterative process. Notwithstanding civil engineering requirements, technical and functional requirements for the processes housed in the buildings, the internal layout must also take into account safety requirements and workers protection requirements in normal, incidental and accidental conditions. Potential hazards and constraints result in the definition of various zonings. Each room is classified depending on the level of the risks induced by normal and incidental or accidental conditions: - radiological exposure to tritium, and activated products → radiological zoning; - exposure to beryllium (vacuum vessel blankets) → beryllium zoning; - exposure to electromagnetic fields during plasma shots → electromagnetic zoning; - fire hazard: fire compartmentalization; ·explosion hazard: anti-blast zoning; - waste treatment → waste zoning. Prevention of common mode failure due to internal hazards (fire, flooding, load drop...) for redundant safety important components is also taken into account by geographical separation or by installation of protections between components (either in the same room or preferably when possible by segregation in different rooms). The designer has also taken into account access requirements to equipments and circulation rules for personal, material and substances in ITER nuclear buildings. Risk analyses are performed in the Preliminary Safety Report in order to demonstrate that the layout of ITER nuclear buildings allows meeting the safety objectives and takes into account the requirements associated with the zonings classifications and complies with the regulation. When necessary, modifications are implemented in the design. For example, recent layout modifications are been studied in the tritium building: utility rooms (electrical supply) were transferred outside the radiological hazard zones. A direct access from

Design issues and cost implications of RTO/RC-ITER divertor

International Nuclear Information System (INIS)

Ibbott, C.; Antipenkov, A.; Chiocchio, S.; Federici, G.; Heidl, H.; Janeschitz, G.; Martin, E.; Tivey, R.

2000-01-01

This paper reports on the conceptual divertor design developed for the reduced technical objectives/reduced cost-international thermonuclear experimental reactor (RTO/RC-ITER). The cost drivers are discussed and a number of cost-reducing measures identified. Scaled costs, based on industrial estimates of the 1998 ITER design (Technical Basis for the ITER Final Design Report, Cost Review and Safety Analysis (FDR). ITER EDA Documentation Series No. 16. IAEA, Vienna, 1998), give for the RTO/RC-ITER ∼60% of the FDR costs. Plasma facing components (PFCs) account for 75% of the total divertor costs. Hence, PFC design simplifications are outlined in the paper showing the possibility of achieving a cost reduction of 50%. The design proposals, outlined in the paper, focus on minimising the number of sub-components and simplifying the manufacturing cycle. These changes contribute to improved reliability based on a more robust coolant design layout. The reduced space allocated to the divertor (G. Janeschitz, A. Antipenkov, V. Barabash, S. Chiocchio, G. Federici, C. Ibbott, E. Martin, R. Tivey, Overview of the Divertor Design and its Integration into RTO/RC-ITER, this conference) requires changes to the design that minimise the cassette body thickness, relocate the cassette attachments and revise the remote handling philosophy. Results of supporting electro-magnetic, neutron shielding, thermo-hydraulic and pumping conductance analyses are reported, qualifying the cassette design. A reduction in the coolant inlet temperature to 100-120 deg. C is discussed in terms of thermal-hydraulic performance and fatigue life of the heat sink. Finally, an R and D plan sets out the work needed: (1) to develop the cost saving measures of the new design; and (2) to demonstrate the reliability of the chosen technologies

Critical Assessment of Pressure Gauges for ITER

International Nuclear Information System (INIS)

Tabares, Francisco L.; Tafalla, David; Garcia-Cortes, Isabel

2008-01-01

The density and flux of molecular species in ITER, largely dominated by the molecular form of the main plasma components and the He ash, is a valuable parameter of relevance not only for operation purposes but also for validating existing neutral particle models of direct implications in divertor performance. An accurate and spatially resolved monitoring of this parameter implies the proper selection of pressure gauges able to cope with the very unique and aggressive environment to be expected in a fusion reactor. To date, there is no standard gauge fulfilling all the requirements, which encompass high neutron and gamma fluxes, together with strong magnetic field and temperature excursions and dusty environment. In the present work, a review of the challenges to face in the measurement of neutral pressure in ITER, together with existing technologies and developments to be made in some of them for their application to the task is presented. Particular attention is paid to R and D needs of existing concepts with potential use in future designs

Operation and control of ITER plasmas

International Nuclear Information System (INIS)

2001-01-01

Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

Operation and control of ITER plasmas

International Nuclear Information System (INIS)

1999-01-01

Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

Design of ITER divertor VUV spectrometer and prototype test at KSTAR tokamak

Science.gov (United States)

Seon, Changrae; Hong, Joohwan; Song, Inwoo; Jang, Juhyeok; Lee, Hyeonyong; An, Younghwa; Kim, Bosung; Jeon, Taemin; Park, Jaesun; Choe, Wonho; Lee, Hyeongon; Pak, Sunil; Cheon, MunSeong; Choi, Jihyeon; Kim, Hyeonseok; Biel, Wolfgang; Bernascolle, Philippe; Barnsley, Robin; O'Mullane, Martin

2017-12-01

Design and development of the ITER divertor VUV spectrometer have been performed from the year 1998, and it is planned to be installed in the year 2027. Currently, the design of the ITER divertor VUV spectrometer is in the phase of detail design. It is optimized for monitoring of chord-integrated VUV signals from divertor plasmas, chosen to contain representative lines emission from the tungsten as the divertor material, and other impurities. Impurity emission from overall divertor plasmas is collimated through the relay optics onto the entrance slit of a VUV spectrometer with working wavelength range of 14.6-32 nm. To validate the design of the ITER divertor VUV spectrometer, two sets of VUV spectrometers have been developed and tested at KSTAR tokamak. One set of spectrometer without the field mirror employs a survey spectrometer with the wavelength ranging from 14.6 nm to 32 nm, and it provides the same optical specification as the spectrometer part of the ITER divertor VUV spectrometer system. The other spectrometer with the wavelength range of 5-25 nm consists of a commercial spectrometer with a concave grating, and the relay mirrors with the same geometry as the relay mirrors of the ITER divertor VUV spectrometer. From test of these prototypes, alignment method using backward laser illumination could be verified. To validate the feasibility of tungsten emission measurement, furthermore, the tungsten powder was injected in KSTAR plasmas, and the preliminary result could be obtained successfully with regard to the evaluation of photon throughput. Contribution to the Topical Issue "Atomic and Molecular Data and their Applications", edited by Gordon W.F. Drake, Jung-Sik Yoon, Daiji Kato, Grzegorz Karwasz.

Preliminary Geochemical and Rock Magnetic Study of a Stalagmite From Quintana Roo, Northeastern Yucatan Peninsula

Science.gov (United States)

Urrutia-Fucugauchi, J.; Perez-Cruz, L.; Zhao, X.; Rebolledo-Vieyra, M.; Rodriguez, A.

2012-04-01

We present the preliminary results of geochemical, stable isotopes and rock magnetic studies of a stalagmite from a cave in eastern Quintana Roo, northern Yucatan peninsula. In the past years, there has been increased interest in understanding the paleoclimatic and paleoenvironmental evolution of the Yucatan peninsula and northern Central America, investigating the relationships between climate variations and the development of the Maya civilization. In particular, the variations in regional precipitation and occurrence of several drought periods, which might have been related to the collapse of the Classic Maya period. Stable isotope data on speleothems from different sites in Yucatan and Central America have provided evidence on changes in precipitation, which have affected the Maya region. The stalagmite is ~47 cm long and about 4-5 cm wide at its base. It was collected from the Hilariós Well cave in Tulum, Quintana Roo. Magnetic susceptibility and geochemical analyses have been completed as part of the initial characterization of the stalagmite, with measurements taken every centimeter. Geochemical analyses have been carried out for x-ray fluorescence, with a Niton XRF analyzer. Magnetic susceptibility was determined with a Bartington MS2 instrument using the high resolution surface probe. Additional rock magnetic analyses include magnetic hysteresis loops and isothermal remanent magnetization (IRM) acquisition, and saturation IRM demagnetization, which have been measured with a MicroMag instrument. Hysteresis loops are diamagnetic, with small varying low-coercivity ferromagnetic components. The elemental compositions of major oxides and trace elements vary with depth. Calcium is the major element and displays a pattern of small amplitude fluctuations with a trend to lower values at the bottom, which are also shown in other elements such as barium. Silica and elements such as titanium and strontium are positively correlated and show an apparent cyclic pattern

Electromagnetic analysis of ITER shield blanket under VDE

International Nuclear Information System (INIS)

Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

2010-01-01

Electromagnetic force and torque of ITER shield blanket system and their surrounding major component under vertical displacement event (VDE) were calculated with finite element method. ANSYS APDL was used to simulate the shape and magnitude of plasmas current dynamically in the VDE course, and external magnetic field was imposed, then the induced current distribution inside the all conductor including the blanket was obtained from the calculation. The force and torque for every blanket module was obtained to assess the safety of blanket system under VDE. (authors)

ITER radio frequency systems

International Nuclear Information System (INIS)

Bosia, G.

1998-01-01

Neutral Beam Injection and RF heating are two of the methods for heating and current drive in ITER. The three ITER RF systems, which have been developed during the EDA, offer several complementary services and are able to fulfil ITER operational requirements

ITER blanket module connectors. Design, analysis and testing for procurement arrangement

International Nuclear Information System (INIS)

Khomiakov, S.; Poddubnyi, I.; Kolganov, V.; Zhmakin, A.; Parshutin, E.; Danilov, I.; Strebkov, Yu.; Skladnov, K.; Vlasov, D.; Cheburova, A.; Romannikov, A.; Raffray, R.; Egorov, K.; Chappuis, Ph.; Sadakov, S.; Calcagno, B.; Roccella, R.

2016-01-01

Highlights: • Procurement Arrangement on Blanket Module Connections (BMC) was signed by ITER Organization and Russian Federation Domestic Agency in late 2014. • “N.A. Dollezhal Research and Development Institute of Power Engineering” (NIKIET) was selected as a general supplier of BMC. • NIKIET plays a key role in design development, analytical and experimental justification and manufacturing of BMC. • NIKIET shall fabricate, test and deliver to ITER 2109 flexible supports, 2561 pads, 1053 electrical straps and 1053 pedestals. - Abstract: A standard ITER Blanket module (BM) is attached to the Vacuum Vessel (VV) with a special system of Blanket Module Connections (BMCs) comprising flexible supports, insulating key pads and electrical straps. BMCs fix the modules relative to the VV and manage the current flow. They accommodate transient, cyclic, thermal and electro-magnetic (EM) loads in a vacuum environment and under neutron radiation. Dynamic, thermal-structural and strength analyses have been performed in support of the BMC design and the results have been experimentally confirmed. The components with uncertain behavior including partially and non-preloaded threads, insulation coating, and electrical contacts were designed by experiments. The effort to develop a reliable and robust design of the BMCs in time for the signature of the Procurement Arrangement on BMCs between ITER Organization and Russian Federation in late 2014 spanned several years. It includes design and analysis as well as experimental activities by the ITER Organization and by JSC “NIKIET” (Russia), which, as an affirmed subcontractor will manufacture and supply BMCs to the ITER site. This paper summarizes the overall effort focusing in particular on the more recent PA supporting activities.

ITER blanket module connectors. Design, analysis and testing for procurement arrangement

Energy Technology Data Exchange (ETDEWEB)

Khomiakov, S., E-mail: khomias58@mail.ru [Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Str. 2/8, Moscow (Russian Federation); Poddubnyi, I.; Kolganov, V.; Zhmakin, A.; Parshutin, E.; Danilov, I.; Strebkov, Yu.; Skladnov, K.; Vlasov, D.; Cheburova, A. [Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Str. 2/8, Moscow (Russian Federation); Romannikov, A. [Institution “Project Center ITER”, 123098, Academic Kurchatov' s Sq.,1, Moscow (Russian Federation); Raffray, R.; Egorov, K.; Chappuis, Ph.; Sadakov, S.; Calcagno, B.; Roccella, R. [ITER Organization, Route de Vinon sur Verdon, 13067 St. Paul-Lez-Durance (France)

2016-11-01

Highlights: • Procurement Arrangement on Blanket Module Connections (BMC) was signed by ITER Organization and Russian Federation Domestic Agency in late 2014. • “N.A. Dollezhal Research and Development Institute of Power Engineering” (NIKIET) was selected as a general supplier of BMC. • NIKIET plays a key role in design development, analytical and experimental justification and manufacturing of BMC. • NIKIET shall fabricate, test and deliver to ITER 2109 flexible supports, 2561 pads, 1053 electrical straps and 1053 pedestals. - Abstract: A standard ITER Blanket module (BM) is attached to the Vacuum Vessel (VV) with a special system of Blanket Module Connections (BMCs) comprising flexible supports, insulating key pads and electrical straps. BMCs fix the modules relative to the VV and manage the current flow. They accommodate transient, cyclic, thermal and electro-magnetic (EM) loads in a vacuum environment and under neutron radiation. Dynamic, thermal-structural and strength analyses have been performed in support of the BMC design and the results have been experimentally confirmed. The components with uncertain behavior including partially and non-preloaded threads, insulation coating, and electrical contacts were designed by experiments. The effort to develop a reliable and robust design of the BMCs in time for the signature of the Procurement Arrangement on BMCs between ITER Organization and Russian Federation in late 2014 spanned several years. It includes design and analysis as well as experimental activities by the ITER Organization and by JSC “NIKIET” (Russia), which, as an affirmed subcontractor will manufacture and supply BMCs to the ITER site. This paper summarizes the overall effort focusing in particular on the more recent PA supporting activities.

Preliminary results of the use of photon-magnetic therapy in prevention and treatment of skin radiation reactions of patients with breast cancer with adjuvant radiation therapy

International Nuclear Information System (INIS)

Syimonova, L.Yi.; Gertman, V.Z.; Byilogurova, L.V.; Kulyinyich, G.V.; Lavrik, V.P.

2012-01-01

The authors report preliminary findings of the investigation of the effect of combination photon-magnetic therapy with successive application of red and blue light to the skin of breast cancer patients during the course of post-operative radiation therapy. It was established that photonmagnetic therapy positively influenced the state of the skin in the irradiated areas. Addition of the magnetic factor significantly improved the efficacy of phototherapy. The patients receiving photon-magnetic therapy finished the course of radiation therapy with almost unchanged skin.

ITER Construction--Plant System Integration

International Nuclear Information System (INIS)

Tada, E.; Matsuda, S.

2009-01-01

This brief paper introduces how the ITER will be built in the international collaboration. The ITER Organization plays a central role in constructing ITER and leading it into operation. Since most of the ITER components are to be provided in-kind from the member countries, integral project management should be scoped in advance of real work. Those include design, procurement, system assembly, testing, licensing and commissioning of ITER.

Magnet systems for the International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Henning, C.D.; Miller, J.R.

1988-01-01

The definition phase for the International Thermonuclear Experimental Reactor (ITER) has been nearly completed, thus beginning a three-year design effort by teams from the European Community (EC), Japan, US, and USSR. Preliminary parameters for the superconducting magnet system have been established to guide more detailed design work. Radiation tolerance of the superconductors and insulators has been important because it sets requirements for the neutron-shield dimension and sensitively influences reactor size. Major levels of mechanical stress appear in the structural cases of the inboard legs of the toroidal-field (TF) coils. The winding packs of the TF coils include significant fractions of steel that provide support against in-plane separating loads, but they offer little support against out-of-plane loads unless shear-bonding of the conductors can be maintained. Heat removal from nuclear and ac loads has not limited the fundamental design, but it has nonnegligible economic consequences. 3 refs., 3 figs., 5 tabs

ITER definition phase

International Nuclear Information System (INIS)

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is envisioned as a fusion device which would demonstrate the scientific and technological feasibility of fusion power. As a first step towards achieving this goal, the European Community, Japan, the Soviet Union, and the United States of America have entered into joint conceptual design activities under the auspices of the International Atomic Energy Agency. A brief summary of the Definition Phase of ITER activities is contained in this report. Included in this report are the background, objectives, organization, definition phase activities, and research and development plan of this endeavor in international scientific collaboration. A more extended technical summary is contained in the two-volume report, ''ITER Concept Definition,'' IAEA/ITER/DS/3. 2 figs, 2 tabs

United States rejoin ITER

International Nuclear Information System (INIS)

Roberts, M.

2003-01-01

Upon pressure from the United States Congress, the US Department of Energy had to withdraw from further American participation in the ITER Engineering Design Activities after the end of its commitment to the EDA in July 1998. In the years since that time, changes have taken place in both the ITER activity and the US fusion community's position on burning plasma physics. Reflecting the interest in the United States in pursuing burning plasma physics, the DOE's Office of Science commissioned three studies as part of its examination of the option of entering the Negotiations on the Agreement on the Establishment of the International Fusion Energy Organization for the Joint Implementation of the ITER Project. These were a National Academy Review Panel Report supporting the burning plasma mission; a Fusion Energy Sciences Advisory Committee (FESAC) report confirming the role of ITER in achieving fusion power production, and The Lehman Review of the ITER project costing and project management processes (for the latter one, see ITER CTA Newsletter, no. 15, December 2002). All three studies have endorsed the US return to the ITER activities. This historical decision was announced by DOE Secretary Abraham during his remarks to employees of the Department's Princeton Plasma Physics Laboratory. The United States will be working with the other Participants in the ITER Negotiations on the Agreement and is preparing to participate in the ITA

Preliminary investigation on welding and cutting methods for first wall support leg in ITER blanket module

International Nuclear Information System (INIS)

Mohri, Kensuke; Suzuki, Satoshi; Enoeda, Mikio; Kakudate, Satoshi; Shibanuma, Kiyoshi; Akiba, Masato

2006-08-01

Concept of a module type of blanket has been applied to ITER shield blanket, of which size is typically 1mW x 1mH x 0.4mB with the weight of 4 ton, in order to enhance its maintainability and fabricability. Each shield blanket module consists of a shield block and four first walls which are separable from the shield block for the purpose of reduction of an electro-magnetic force in disruption events, radio-active waste reduction in the maintenance work and cost reduction in fabrication process. A first wall support leg, a part of the first wall component located between the first wall and the shield block, is required not only to be connected metallurgically to the shield block in order to withstand the electro-magnetic force and coolant pressure, but also to be able to replace the first wall more than 2 times in the hot cell during the life time of the reactor. Therefore, the consistent structure where remote handling equipment can be access to the joint and carry out the welding/cutting works perfectly to replace the first wall in the hot cell is required in the shield blanket design. This study shows an investigation of the blanket module no.10 design with a new type of the first wall support leg structure based on Disc-Cutter technology, which had been developed for the main pipe cutting in the maintenance phase and was selected out of a number of candidate methods, taking its large advantages into account, such as 1) a post-treatment can be eliminated in the hot cell because of no making material chips and of no need of lubricant, 2) the cut surface can be rewelded without any machining. And also, a design for the small type of Disc-Cutter applied to the new blanket module no.10 has been investigated. In conclusion, not only the good performance of Disc-Cutter technology applied to the updated blanket module, but also consistent structure of the simplified shield blanket module including the first wall support leg in order to satisfy the requirements in the

ITER towards the construction

International Nuclear Information System (INIS)

Shimomura, Y.

2005-01-01

The ITER Project has been significantly developed in the last few years in preparation for its construction. The ITER Participant's Negotiators have developed the Joint Implementation Agreement (JIA), ready for finalisation following selection of the construction site and nomination of the project's Director General. The ITER International Team and Participant Teams have continued technical and organisational preparations. Construction will be able to start immediately after the international ITER organisation is established, following signature of the JIA. The Project is strongly supported by the governments of the Participants as well as by the scientific community. The real negotiations, including siting and the final details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. Their wish to host ITER construction is too strong to allow convergence to a single site considering the ITER device in isolation. A broader collaboration among the Parties is therefore being contemplated, covering complementary activities to help accelerate fusion development towards a viable power source, and allow the Participants to reach a conclusion on ITER siting. This report reviews these preparations, and the status of negotiations

Experimental investigation on streaming due to a gap between blanket modules in ITER

International Nuclear Information System (INIS)

Konno, Chikara; Maekawa, Fujio; Oyama, Yukio; Uno, Yoshitomo; Kasugai, Yoshimi; Maekawa, Hiroshi; Ikeda, Yujiro; Wada, Masayuki

2000-01-01

A gap streaming experiment was performed by using a D-T neutron source at FNS/JAERI as the ITER/EDA R and D Task T-218, in order to examine the streaming effects due to gap between shield blanket modules in ITER. The experiment had three phases. The first one defined neutron source characteristics (Source Characterization Experiment), the second (Experiment-l ) aimed at shield for welding part between shield blanket and back plate and the third (Experiment-2) focused on the influence that the gap between shield blanket modules would have on superconducting magnet. The effects of gap streaming were examined in detail experimentally. (author)

ITER-FEAT operation

International Nuclear Information System (INIS)

Shimomura, Y.; Huguet, M.; Mizoguchi, T.; Murakami, Y.; Polevoi, A.R.; Shimada, M.; Aymar, R.; Chuyanov, V.A.; Matsumoto, H.

2001-01-01

ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first ten years of operation will be devoted primarily to physics issues at low neutron fluence and the following ten years of operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes, such as inductive high Q modes, long pulse hybrid modes and non-inductive steady state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours a day but also in involving the worldwide fusion community and in promoting scientific competition among the ITER Parties. (author)

Lumbar muscle activity during common lifts: a preliminary study using magnetic resonance imaging.

Science.gov (United States)

Mayer, John M; Graves, James E; Manini, Todd M; Nuzzo, James L; Ploutz-Snyder, Lori L

2013-04-01

The purpose of this preliminary study was to assess lumbar multifidus, erector spinae, and quadratus lumborum muscle activity during lifts as measured by changes in transverse relaxation time (T2) from magnetic resonance imaging (MRI). Thirteen healthy adults performed dynamic squat, stoop, and asymmetric stoop lifts at a standard load, with each lift followed by MRI. Increase in T2 for the multifidus and erector spinae was greater for the stoop than squat. No difference in T2 increase was noted between the multifidus and erector spinae for the squat or stoop. Increase in T2 for the contralateral multifidus was less for the asymmetric stoop than stoop. Future research using MRI and other biomechanical techniques is needed to fully characterize lumbar muscle activity during lifts for various populations, settings, postures, and loads.

The European Fusion Research and Development Programme and the ITER Project

International Nuclear Information System (INIS)

Green, B J

2006-01-01

The EURATOM fusion research and development programme is a well integrated and coordinated programme. It has the objective of ''developing the technology for a safe, sustainable, environmentally responsible and economically viable energy source.'' The programme is focussed on the magnetic confinement approach and supports 23 Associations which involve research entities (many with experimental and technology facilities) each having a bilateral contractual relationship with the European Commission. The paper will describe fusion reactions and present their potential advantages as an energy source. Further, it will describe the EURATOM programme and how it is organised and implemented. The success of the European programme and that of other national programmes, have provided the basis for the international ITER Project, which is the next logical step in the development of fusion energy. The paper will describe ITER, its aims, its design, and the supporting manufacture of prototype components. The European contribution to ITER, the exploitation of the Joint European Torus (JET), and the long-term reactor technology R and D are carried out under the multilateral European Fusion Development Agreement (EFDA)

The European fusion research and development programme and the ITER Project

International Nuclear Information System (INIS)

Green, B.J.

2004-01-01

The EURATOM fusion R and D programme is a well integrated and co-ordinated programme a good example of a European Research Area. Its goal is 'the joint creation of prototype reactors for power stations to meet the needs of society: operational safety, environmental compatibility, economic viability'. The programme is focussed on the magnetic confinement approach to fusion energy and supports 21 associated laboratories and a range of experimental and fusion technology facilities. The paper will briefly describe this programme and how it is organised and implemented. Its success and that of other national programmes has defined the international ITER Project, which is the next logical step in fusion R and D. The paper will describe ITER, its aims, its design, and the supporting manufacture of prototype components. The European contribution to ITER, as well as the exploitation of the Joint European Torus (JET) and long-term fusion reactor technology R and D are carried out under the European Fusion Development Agreement (EFDA). Finally, the potential advantages of fusion as an energy source will be presented. (author)

ITER CTA newsletter. No. 2

International Nuclear Information System (INIS)

2001-10-01

This ITER CTA newsletter contains results of the ITER toroidal field model coil project presented by ITER EU Home Team (Garching) and an article in commemoration of the late Dr. Charles Maisonnier, one of the former leaders of ITER who made significant contributions to its development

Progress on ITER remote experimentation centre

International Nuclear Information System (INIS)

Ozeki, Takahisa; Clement-Lorenzo, Susana; Nakajima, Noriyoshi

2016-01-01

Construction of ITER remote experimentation centre (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is progressing. In order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, development of a remote experiment system by using the Satellite Tokamak (JT-60SA) facilities was planned and the development of software for the remote experiment is ongoing, including the systems for the remote connection and the communication between the remote site and the on-site facility. The network system from REC in Rokkasho-site of Japan to the network in EU was established in collaboration with the National Institute of Informatics (NII). Effective data transfer method that is capable of fast transfer speeds in the gigabit range is investigated. Data transfer at the rate of several Gbps was successfully obtained between the institutes in Japan. The preliminary versions of the software for data analysis are developed, such as for visualization of time dependent experimental data and transport simulations, visualization of plasma boundary/equilibrium and spatial profiles of diagnostic data. The remote data access program and an integrated platform for Documentation and Experiment Management are also being developed. A remote experiment room in the Rokkasho-site in Japan was designed and the construction started. The function of REC will be tested and the total system will be demonstrated by the middle of 2017.

Progress on ITER remote experimentation centre

Energy Technology Data Exchange (ETDEWEB)

Ozeki, Takahisa, E-mail: ozeki.takahisa@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Obuchi Rokkasho, Kitakami-gun, Aomori 039-3212 (Japan); Clement-Lorenzo, Susana [Fusion for Energy, Torres Diagonal Litoral, B3, 13/03, Barcelona 08019 (Spain); Nakajima, Noriyoshi [National institute for Fusion Science and Project leader of IFERC, 2-166 Obuchi, Rokkasho, Kamikita-gun, Aomori 039-3212 (Japan)

2016-11-15

Construction of ITER remote experimentation centre (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is progressing. In order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, development of a remote experiment system by using the Satellite Tokamak (JT-60SA) facilities was planned and the development of software for the remote experiment is ongoing, including the systems for the remote connection and the communication between the remote site and the on-site facility. The network system from REC in Rokkasho-site of Japan to the network in EU was established in collaboration with the National Institute of Informatics (NII). Effective data transfer method that is capable of fast transfer speeds in the gigabit range is investigated. Data transfer at the rate of several Gbps was successfully obtained between the institutes in Japan. The preliminary versions of the software for data analysis are developed, such as for visualization of time dependent experimental data and transport simulations, visualization of plasma boundary/equilibrium and spatial profiles of diagnostic data. The remote data access program and an integrated platform for Documentation and Experiment Management are also being developed. A remote experiment room in the Rokkasho-site in Japan was designed and the construction started. The function of REC will be tested and the total system will be demonstrated by the middle of 2017.

Analysis of ITER NbTi and Nb3Sn CICCs experimental minimum quench energy with JackPot, MCM and THEA models

NARCIS (Netherlands)

Bagni, T.; Duchateau, J.L.; Breschi, M.; Devred, A.; Nijhuis, A.

2017-01-01

Cable-in-conduit conductors (CICCs) for ITER magnets are subjected to fast changing magnetic fields during the plasma-operating scenario. In order to anticipate the limitations of conductors under the foreseen operating conditions, it is essential to have a better understanding of the stability

Piping structural design for the ITER thermal shield manifold

Energy Technology Data Exchange (ETDEWEB)

Noh, Chang Hyun, E-mail: chnoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Chung, Wooho, E-mail: whchung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo; Kang, Kyoung-O. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bae, Jing Do; Cha, Jong Kook [Korea Marine Equipment Research Institute, Busan 606-806 (Korea, Republic of); Kim, Kyoung-Kyu [Mecha T& S, Jinju-si 660-843 (Korea, Republic of); Hamlyn-Harris, Craig; Hicks, Robby; Her, Namil; Jun, Chang-Hoon [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

2015-10-15

Highlights: • We finalized piping design of ITER thermal shield manifold for procurement. • Support span is determined by stress and deflection limitation. • SQP, which is design optimization method, is used for the pipe design. • Benchmark analysis is performed to verify the analysis software. • Pipe design is verified by structural analyses. - Abstract: The thermal shield (TS) provides the thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2 K. The TS is actively cooled by 80 K pressurized helium gas which flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the manifold piping design and analysis for the ITER thermal shield. First, maximum allowable span for the manifold support is calculated based on the simple beam theory. In order to accommodate the thermal contraction in the manifold feeder, a contraction loop is designed and applied. Sequential Quadratic Programming (SQP) method is used to determine the optimized dimensions of the contraction loop to ensure adequate flexibility of manifold pipe. Global structural behavior of the manifold is investigated when the thermal movement of the redundant (un-cooled) pipe is large.

The design, magnetization and control of a superconducting permanent magnet synchronous motor

Energy Technology Data Exchange (ETDEWEB)

Jiang, Y; Pei, R; Xian, W; Hong, Z; Coombs, T A [Engineering Department, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom)], E-mail: yj222@cam.ac.uk

2008-06-15

This paper describes in detail the method of magnetization of a superconducting permanent magnet synchronous motor. The rotor of the motor consists of 60 superconducting pucks, which are magnetized by two additional copper windings. The pulse field magnetization (PFM) method is considered and the resulted distribution of the magnetizing flux linkage from the rotor is not a perfect sine wave in the air gap, which leads to a large torque ripple and harmonics of the stator currents. In order to suppress the torque ripple, an iterative learning control (ILC) algorithm is used in addition to the former field-oriented control method. The results show the ILC algorithm can largely reduce the torque ripple.

Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

Energy Technology Data Exchange (ETDEWEB)

Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

2014-08-21

In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

Science.gov (United States)

Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

2014-08-01

In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

International Nuclear Information System (INIS)

Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

2014-01-01

In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed

Structural analysis for the joint of the ITER ELM coil

Energy Technology Data Exchange (ETDEWEB)

Zhang, Shanwen, E-mail: zhangshanwen123@163.com [College of Mechanical Engineering Yangzhou University, Yangzhou 225127 (China); Song, Yuntao; Wang, Zhongwei; Ji, Xiang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 200031 (China); Zhang, Jianfeng [College of Mechanical Engineering Yangzhou University, Yangzhou 225127 (China)

2017-01-15

Highlights: • The FE sub-model method is feasible and rapid for the joint design. • The components meet the static and fatigue criteria. • Nuclear heat is the key factor for the joint design. - Abstract: The joint is an important component of the Edge Localized Modes (ELM) coils in fusion reactor, which is used to connect the ELM coils. Like the ELM coils, the joints work in an environment with high radiation levels, high temperature and high magnetic field. These joints are mainly subject to nuclear heat from the plasma and cyclic electromagnetic (EM) loads induced by the interaction of ELM coil current with magnetic fields. Take the joint of ITER ELM coil for example. In order to assure the structural integrity of joints under these loads, it is necessary to estimate the strength and fatigue of the joints. As a local model, the joint without ELM coil is investigated by the sub-model method. Results show that the finite element sub-model method is feasible and rapid for the joint design. The maximum magnetic flux intensity occurs in the axial direction for the joint local reference, which parallels with the current and corresponds to the toroidal direction of the ITER. The two areas at the top of the Inconel sleeve appear high temperature. For the joint, the conductor, jacket and sleeve can meet the static and fatigue criteria and the joint design is valid and feasible. The thermal load from the nuclear heat is the key factor for the joint design.

Computational models for electromagnetic transients in ITER vacuum vessel, cryostat and thermal shield

International Nuclear Information System (INIS)

Alekseev, A.; Arslanova, D.; Belov, A.; Belyakov, V.; Gapionok, E.; Gornikel, I.; Gribov, Y.; Ioki, K.; Kukhtin, V.; Lamzin, E.; Sugihara, M.; Sychevsky, S.; Terasawa, A.; Utin, Y.

2013-01-01

A set of detailed computational models are reviewed that covers integrally the system “vacuum vessel (VV), cryostat, and thermal shields (TS)” to study transient electromagnetics (EMs) in the ITER machine. The models have been developed in the course of activities requested and supervised by the ITER Organization. EM analysis is enabled for all ITER operational scenarios. The input data are derived from results of DINA code simulations. The external EM fields are modeled accurate to the input data description. The known magnetic shell approach can be effectively applied to simulate thin-walled structures of the ITER machine. Using an integral–differential formulation, a single unknown is determined within the shells in terms of the vector electric potential taken only at the nodes of a finite-element (FE) mesh of the conducting structures. As a result, the FE mesh encompasses only the system “VV + Cryostat + TS”. The 3D model requires much higher computational resources as compared to a shell model based on the equivalent approximation. The shell models have been developed for all principal conducting structures in the system “VV + Cryostat + TS” including regular ports and neutral beam ports. The structures are described in details in accordance with the latest design. The models have also been applied for simulations of EM transients in components of diagnostic systems and cryopumps and estimation of the 3D effects of the ITER structures on the plasma performance. The developed models have been elaborated and applied for the last 15 years to support the ITER design activities. The finalization of the ITER VV design enables this set of models to be considered ready to use in plasma-physics computations and the development of ITER simulators

Detailed 3-D nuclear analysis of ITER outboard blanket modules

Energy Technology Data Exchange (ETDEWEB)

Bohm, Tim, E-mail: tdbohm@wisc.edu [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, Michael; Bullock, James [Formerly, Fusion Technology, Sandia National Laboratories, Albuquerque, NM (United States)

2015-10-15

Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

Detailed 3-D nuclear analysis of ITER outboard blanket modules

International Nuclear Information System (INIS)

Bohm, Tim; Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul; Ulrickson, Michael; Bullock, James

2015-01-01

Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

HL-2A experiment and ITER-related activity at SWIP

International Nuclear Information System (INIS)

Duan Xuru

2007-01-01

In this overview the recent progress on HL-2A tokamak experiment and ITER-related activity at SWIP is summarized. Experiment on HL-2A is one of the important research activities at SWIP. In the last two years, some new hardware had been developed, these include four sets of ECRH system with a total power up to 2 MW, new diagnostics such as 8-channel laser interferometer. The studied subjects were focused on plasma auxiliary heating, fuelling, transport, edge plasma physics and turbulence, etc. Progress in these fields has been obtained. For example, the toroidal symmetry of the geodesic acoustic mode (GAM), the oscillating branch of zonal flows has been demonstrated for the first time using a novel 3-step Langmuir Probe, and the poloidal and radial structure of the low frequency electric potential and field were simultaneously observed. During ECRH experiments under different discharge conditions, the MHD instability excited by high energetic electrons was investigated. Besides, non-local heat transport due to SMBI during ECRH was studied. Another important fusion activity at SWIP is the ITER relevant technology. The R and D of four ITER procurements (first wall and shielding blanket, magnet gravity support, gas injection and glow discharge cleaning system, neutron flux measurement) has been undertaken. Progress has been made, e.g. the technology for manufacturing high purity (>99%) ITER specified Be plate and CuCrZr alloy is obtained, their major mechanical and physical properties were measured. For ITER-TBM, a structural material named as CLF-1, a type of reduced activation ferritic/martenstic steel, was developed. Besides, some progress in fusion reactor design and related technology was achieved. (authors)

Final design of the Korean AC/DC converters for the ITER coil power supply system

Energy Technology Data Exchange (ETDEWEB)

Oh, Jong-Seok, E-mail: jsoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Choi, Jungwan; Suh, Jae-Hak; Choi, Jihyun [ITER Korea, National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Lee, Lacksang; Kim, Changwoo; Park, Hyungjin; Jo, Seongman; Lee, Seungyun; Hwang, Kwangcheol; Liu, Hyoyol [Dawonsys Corp., Siheung 429-450 (Korea, Republic of); Hong, Ki-Don; Sim, Dong-Joon; Lee, Jang-Soo [Hyosung Corp., Gongdeok-Dong, Seoul 121-720 (Korea, Republic of); Lee, Eui-Jae; Kwon, Yang-Hae; Lee, Dae-Yeol; Ko, Ki-Won; Kim, Jong-Min [Mobiis Corp., Yangjae-dong, Seoul 137-888 (Korea, Republic of); Song, Inho [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); and others

2015-10-15

The final design of the ITER TF, CS, CC and VS AC/DC converters has been completed to implement ITER requirements following the detailed design and refinements of the preliminary design. The number of parallel thyristors and the rating of fuses are coordinated to keep those devices within the explosion limit even under most severe fault conditions. The impedance of the converter transformer has been optimized taking into account the energization inrush current, short circuit current, reactive power consumption and the available DC voltage. To ensure system integrity, AC/DC converters are mechanically divided into transformers, AC busbars, 6-pulse bridges, DC interconnecting busbars and DC reactors, and then all subsystems are decoupled by flexible links. To provide stable real time network communication down to the converters, a one GbE link is deployed between master controllers and local controllers. IEEE 1588 is implemented to the embedded controllers for precision time synchronization. This paper describes the detailed solutions implemented in the final design for the ITER AC/DC converters with R&D results of converter prototypes.

Spirit and prospects of ITER

Energy Technology Data Exchange (ETDEWEB)

Velikhov, E.P. [Kurchatov Institute of Atomic Energy, Moscow (Russian Federation)

2002-10-01

ITER is the unique and the most straightforward way to study the burning plasma science in the nearest future. ITER has a firm physics ground based on the results from the world tokamaks in terms of confinement, stability, heating, current drive, divertor, energetic particle confinement to an extend required in ITER. The flexibility of ITER will allow the exploration of broad operation space of fusion power, beta, pulse length and Q values in various operational scenarios. Success of the engineering R and D programs has demonstrated that all party has an enough capability to produce all the necessary equipment in agreement with the specifications of ITER. The acquired knowledge and technologies in ITER project allow us to demonstrate the scientific and technical feasibility of a fusion reactor. It can be concluded that ITER must be constructed in the nearest future. (author)

Spirit and prospects of ITER

International Nuclear Information System (INIS)

Velikhov, E.P.

2002-01-01

ITER is the unique and the most straightforward way to study the burning plasma science in the nearest future. ITER has a firm physics ground based on the results from the world tokamaks in terms of confinement, stability, heating, current drive, divertor, energetic particle confinement to an extend required in ITER. The flexibility of ITER will allow the exploration of broad operation space of fusion power, beta, pulse length and Q values in various operational scenarios. Success of the engineering R and D programs has demonstrated that all party has an enough capability to produce all the necessary equipment in agreement with the specifications of ITER. The acquired knowledge and technologies in ITER project allow us to demonstrate the scientific and technical feasibility of a fusion reactor. It can be concluded that ITER must be constructed in the nearest future. (author)

Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

Science.gov (United States)

Petrov, S. Ya.; Afanasyev, V. I.; Melnik, A. D.; Mironov, M. I.; Navolotsky, A. S.; Nesenevich, V. G.; Petrov, M. P.; Chernyshev, F. V.; Kedrov, I. V.; Kuzmin, E. G.; Lyublin, B. V.; Kozlovski, S. S.; Mokeev, A. N.

2017-12-01

The control of the deuterium-tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10-200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1-4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

ITER interim design report package documents

International Nuclear Information System (INIS)

1996-01-01

This publication contains the Excerpt from the ITER Council (IC-8), the ITER Interim Design Report, Cost Review and Safety Analysis, ITER Site Requirements and ITER Site Design Assumptions and the Excerpt from the ITER Council (IC-9). 8 figs, 2 tabs

Engineering issues on the diagnostic port integration in ITER upper port 18

Energy Technology Data Exchange (ETDEWEB)

Pak, Sunil, E-mail: paksunil@nfri.re.kr [National Fusion Research Institute, Gwahak-ro, Yuseong-gu, Daejeon (Korea, Republic of); Bertalot, Luciano [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Cheon, Mun Seong [National Fusion Research Institute, Gwahak-ro, Yuseong-gu, Daejeon (Korea, Republic of); Giacomin, Thibaud [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Heemskerk, Cock J.M.; Koning, Jarich F. [Heemskerk Innovative Technology, Merelhof 2, 2172 HZ Sassenheim (Netherlands); Lee, Hyeon Gon [National Fusion Research Institute, Gwahak-ro, Yuseong-gu, Daejeon (Korea, Republic of); Nemtcev, Grigorii [Institution “PROJECT CENTER ITER”, Akademika Kurchatova sq., Moscow (Russian Federation); Ronden, Dennis M.S. [FOM Institute DIFFER, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Seon, Chang Rae [National Fusion Research Institute, Gwahak-ro, Yuseong-gu, Daejeon (Korea, Republic of); Udintsev, Victor; Yukhnov, Nikolay [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Zvonkov, Alexander [Institution “PROJECT CENTER ITER”, Akademika Kurchatova sq., Moscow (Russian Federation)

2016-11-01

Highlights: • Diagnostic port integration in the upper port 18 of ITER is presented in order to house the three diagnostic systems. • Issue on the neutron shielding in the upper port 18 is addressed and the shut-down dose rate in the interspace is summarized. • The maintenance strategy in the upper port 18 is described. - Abstract: The upper port #18 (UP18) in ITER hosts three diagnostic systems: the neutron activation system, the Vacuum Ultra-Violet spectrometer system, and the vertical neutron camera. These diagnostics are integrated into three infrastructures in the port: the upper port plug, interspace support structure and port cell support structure. The port integration in UP18 is at the preliminary design stage and the current design of the infrastructure as well as the diagnostic integration is described here. The engineering issues related to neutron shielding and maintenance are addressed and the design approach is suggested.

ITER CTA newsletter. No. 6

International Nuclear Information System (INIS)

2002-01-01

This ITER CTA Newsletter issue comprises information about the following ITER Meetings: The second negotiation meeting on the joint implementation of ITER, held in Tokyo(Japan) on 22-23 January 2002, and an international ITER symposium on burning plasma science and technology, held the day later after the second negotiation meeting at the same place

Magnetic crustal thickness in Greenland from CHAMP and Ørsted data

DEFF Research Database (Denmark)

Maule, Cathrine Fox; Purucker, Michael E.; Olsen, Nils

2005-01-01

and observatory data. After correcting for the remanent magnetization, we determine the vertically integrated magnetization of the crust. Making some simplifying assumptions about the susceptibility, the thickness of the magnetic crust is determined by iteratively improving an initial crustal thickness model...

A design study of hydrogen isotope separation system for ITER-FEAT

International Nuclear Information System (INIS)

Iwai, Yasunori; Yamanishi, Toshihiko; Nishi, Masataka

2001-03-01

Preliminary design study of the hydrogen isotope separation system (ISS) for the fuel cycle of the ITER-FEAT, a fusion experimental reactor, was carried out based on the substantial reduction of hydrogen flow to the ISS resulting from the design study for scale reduction of the formerly-designed ITER. Three feed streams (plasma exhaust gas stream, streams from the water detritiation system and that from the neutral beam injectors) are fed to the ISS, and three product streams (high purity tritium gas, high purity deuterium gas and hydrogen gas) are made in it by the method of cryogenic distillation. In this study, an original four-column cascade was proposed to the ISS cryogenic distillation column system considering simplification and the operation scenario of the ITER-FEAT. Substantial reduction of tritium inventory in the ISS was found to be possible in the progress of investigation concerning of the corresponding flow rate of tritium product stream (T>90 %) for pellet injector which depends upon the operation condition. And it was found that tritium concentration in the released hydrogen stream into environment from the ISS could easily fluctuate with current design of column arrangement due to the small disturbance in mass flow balance in the ISS. To solve this problem, two-column system for treatment of this flow was proposed. (author)

ITER Status and Plans

Science.gov (United States)

Greenfield, Charles M.

2017-10-01

The US Burning Plasma Organization is pleased to welcome Dr. Bernard Bigot, who will give an update on progress in the ITER Project. Dr. Bigot took over as Director General of the ITER Organization in early 2015 following a distinguished career that included serving as Chairman and CEO of the French Alternative Energies and Atomic Energy Commission and as High Commissioner for ITER in France. During his tenure at ITER the project has moved into high gear, with rapid progress evident on the construction site and preparation of a staged schedule and a research plan leading from where we are today through all the way to full DT operation. In an unprecedented international effort, seven partners ``China, the European Union, India, Japan, Korea, Russia and the United States'' have pooled their financial and scientific resources to build the biggest fusion reactor in history. ITER will open the way to the next step: a demonstration fusion power plant. All DPP attendees are welcome to attend this ITER town meeting.

Theory of neoclassical tearing modes and its application to ITER-FEAT

International Nuclear Information System (INIS)

Pustovitov, V.D.; Mikhailovskii, A.B.; Konovalov, S.V.; Zvonkov, A.V.; Kobayashi, N.; Mukhovatov, V.S.

2001-01-01

Neoclassical tearing modes (NTM) can be responsible for beta limitation in long-pulse ITER discharges. The excitation and growth of NTM are governed by the competing bootstrap current, polarization current and so-called Δ' effects. Also, the magnetic well and Electron Cyclotron Current Drive (ECCD) can stabilize the NTM. We study analytically and numerically all the effects with a particular emphasis on the polarization current in the analytical part of our study. We show that the polarization current description requires a generalized transport theory including the hyperviscosity, electron pressure gradient and, as well, the finite ion Larmor radius effects in the perpendicular current. The profile function nonstationarity must be taken into account for calculation of the island rotation frequency. Results of numerical simulation of NTM suppression by modulated ECCD in ITER are presented. (author)

ITER council proceedings: 1999

International Nuclear Information System (INIS)

1999-01-01

In 1999 the ITER meeting in Cadarache (10-11 March 1999) and the Programme Directors Meeting in Grenoble (28-29 July 1999) took place. Both meetings were exclusively devoted to ITER engineering design activities and their agendas covered all issues important for the development of ITER. This volume presents the documents of these two important meetings

ITER EDA technical activities

International Nuclear Information System (INIS)

Aymar, R.

1998-01-01

Six years of technical work under the ITER EDA Agreement have resulted in a design which constitutes a complete description of the ITER device and of its auxiliary systems and facilities. The ITER Council commented that the Final Design Report provides the first comprehensive design of a fusion reactor based on well established physics and technology

Edge localized modes control by resonant magnetic perturbations

International Nuclear Information System (INIS)

Nardon, E.

2007-10-01

The present work is dedicated to one of the most promising methods of control of the ELMs (Edge Localized Modes), based on a system of coils producing Resonant Magnetic Perturbations (RMPs). Our main objectives are, on the one hand, to improve the physical understanding of the mechanisms at play, and on the other hand to propose a concrete design of ELMs control coils for ITER. In order to calculate and analyze the magnetic perturbations produced by a given set of coils, we have developed the ERGOS code. The first ERGOS calculation was for the DIII-D ELMs control coils, the I-coils. It showed that they produce magnetic islands chains which overlap at the edge of the plasma, resulting in the ergodization of the magnetic field. We have then used ERGOS for the modelling of the experiments on ELMs control using the error field correction coils at JET and MAST. In the case of JET, we have shown the existence of a correlation between the mitigation of the ELMs and the ergodization of the magnetic field at the edge, in agreement with the DIII-D result. In order to design the ELMs control coils for ITER we have used ERGOS intensively, taking the case of the DIII-D I-coils as a reference. Three candidate designs came out, which we presented at the ITER Design Review, in 2007. Recently, the ITER management decided to provide a budget for building ELMs control coils, the design of which remains to be chosen between two of the three options that we proposed. Finally, in order to understand better the non-linear magnetohydrodynamics phenomena taking place in ELMs control by RMPs, we performed numerical simulations, in particular with the JOREK code for a DIII-D case. The simulations reveal the existence of convection cells induced at the edge by the magnetic perturbations, and the possible screening of the RMPs in presence of rotation

Impact of risk analysis on ITER nuclear buildings internal layout

International Nuclear Information System (INIS)

Lignini, F.; Sabathe, L.; Rodriguez-Rodrigo, L.; Uzan-Elbez, J.; Girard, J.-P.

2006-01-01

The internal layout of ITER nuclear buildings (Tokamak, Tritium, Hot cell and Radwaste buildings) is the result of an iterative process. Notwithstanding civil engineering requirements, technical and functional requirements for the processes housed in the buildings, the internal layout must also take into account safety requirements and workers protection requirements in normal, incidental and accidental conditions. Potential hazards and constraints result in the definition of various zonings. Each room is classified depending on the level of risk induced by normal and incidental or accidental conditions : - radiological exposure to tritium, and activated products → radiological zoning - exposure to beryllium (vacuum vessel blankets) → beryllium zoning - exposure to electromagnetic fields during plasma shots → electromagnetic zoning - fire hazard : fire compartmentalisation - explosion hazard → anti-blast zoning - waste treatment → waste zoning Prevention of common mode failure due to internal hazards (fire, flooding, load drop ...) for redundant safety important components is also taken into account by geographical separation or by installation of protections between components (either in the same room or preferably when possible by segregation in different rooms). The designer also has to take into account access requirements to equipment and circulation rules for personal, material and substances in ITER nuclear buildings. Risk analyses are performed in the Preliminary Safety Report in order to demonstrate that the layout of ITER nuclear buildings allows to meet the safety objectives and takes into account the requirements associated with the zonings classifications and complies with the regulation. When necessary, modifications are implemented in the design. For example, recent layout modifications were included in the tritium building : utility rooms (electrical supply) were transferred outside the radiological hazard zones. A direct access from the

Future plan of ITER

International Nuclear Information System (INIS)

Kitsunezaki, Akio

1998-01-01

In cooperation of four countries, Japan, USA, EU and Russia, ITER plan has been proceeding as ''the conceptual design activities'' from 1988 to 1990 and ''the industrial design activities'' since 1992. To construct ITER, the legal and work side of ITER operation has been investigated by four countries. However, their economic conditions have been changed to be wrong. So that, construction of ITER can not begin after end of industrial design activities in 1998. Accordingly, they determined to continue the industrial design activities more three years in order to study low cost options and to test the superconductive model·coil. (S.Y.)

ITER council proceedings: 1992

International Nuclear Information System (INIS)

1994-01-01

At the signing of the ITER EDA Agreement on July, 1992, each of the Parties presented to the Director General the names of their designated members of the ITER Council. Upon receiving those names, the Director General stated that the ITER Engineering Design Activities were ''ready to begin''. The next step in this process was the convening of the first meeting of the ITER Council. The first meeting of the Council, held in Vienna, was opened by Director General Hans Blix. The second meeting was held in Moscow, the formal seat of the Council. This volume presents records of these first two Council meetings and, together with the previous volumes on the text of the Agreement and Protocol 1 and the preparations for their signing respectively, represents essential information on the evolution of the ITER EDA

Runaway electrons beams in ITER disruptions

International Nuclear Information System (INIS)

Fleischmann, H.H.

1993-01-01

In agreement with the initial projections, the potential generation of runaway beams in disruptions of ITER discharges was performed. This analysis was based on the best-available present projections of plasma parameters existing in large-tokamak disruptions. Using these parameters, the potential contributions from various basic mechanisms for the generation of runway electrons were estimated. The envisioned mechanisms included (i) the well-known Dreicer process (assuming an evaporation of the runways from the thermal distribution), (ii) the seeding of runaway beams resulting from the potential presence of trapped high-temperature electrons from the original discharge still remaining in the disruption plasma at time of reclosure of the magnetic surfaces, and (iii) the generation of runaway beams through avalanche exponentiation of low-level seed runaways resulting via close collisions of existing runaways with cold plasma electrons. Finally, the prospective behavior of the any generated runaway beams -- in particular during their decay -- as well as their potential avoidance and/or damage controlled extraction through the use of magnetic perturbation fields also was considered in some detail

Overview of JET results in support of the ITER physics basis

International Nuclear Information System (INIS)

Gormezano, C.

2001-01-01

The JET experimental campaign has focused on studies in support of the ITER physics basis. An overview of the results obtained is given both for the reference ITER scenario, the ELMy H-mode, and for advanced scenarios which in JET are based on Internal Transport Barriers. JET studies for the ELMy H-mode have been instrumental for the definition of ITER-FEAT. Positive elongation and current scaling in the ITER scaling law have been confirmed, but the observed density scaling fits better a two term (core and edge) model. Significant progress in neo-classical tearing mode limits has been made showing that ITER operation seems to be optimised. Effective helium pumping and divertor enrichment is found to be well within ITER requirements. Target asymmetries and H-isotope retention are well simulated by modelling codes taking into account drift flows in the scrape-off plasmas. Striking improvements in fuelling effectiveness have been found with the new high field pellet launch facility. Good progress has been made on scenarios for achieving good confinement at high densities, both with RI modes and with high field side pellets. Significant development of advanced scenarios in view of their application to ITER has been achieved. Integrated advanced scenarios are in good progress with edge pressure control (impurity radiation). An access domain has been explored showing in particular that the power threshold increases with magnetic field but can be significantly reduced when Lower Hybrid current drive is used to produce target plasma with negative shear. The role of ion pressure peaking on MHD has been well documented. Lack of sufficient additional heating power and interaction with the septum at high beta prevents assessment of beta limits (steady plasmas achieved with β N up to 2.6). Plasmas with non-inductive current (I NI /Ip=60%), well aligned with plasma current, high beta and good confinement have also been obtained. (author)

ITER physics design guidelines: 1989

International Nuclear Information System (INIS)

Uckan, N.A.

1990-01-01

The physics basis for ITER has been developed from an assessment of the results of the last twenty-five years of tokamak research and from detailed analysis of important physics issues specifically for the ITER design. This assessment has been carried out with direct participation of members of the experimental teams of each of the major tokamaks in the world fusion program through participation in ITER workshops, contributions to the ITER Physics R and D Program, and by direct contacts between the ITER team and the cognizant experimentalists. Extrapolations to the present data base, where needed, are made in the most cautious way consistent with engineering constraints and performance goals of the ITER. In cases where a working assumptions had to be introduced, which is insufficiently supported by the present data base, is explicitly stated. While a strong emphasis has been placed on the physics credibility of the design, the guidelines also take into account that ITER should be designed to be able to take advantage of potential improvements in tokamak physics that may occur before and during the operation of ITER. (author). 33 refs

ITER council proceedings: 1996

International Nuclear Information System (INIS)

1997-01-01

Records of the 10. ITER Council Meeting (IC-10), held on 26-27 July 1996, in St. Petersburg, Russia, and the 11. ITER Council Meeting (IC-11) held on 17-18 December 1996, in Tokyo, Japan, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the cost review and safety analysis. Figs, tabs

The ITER EC H&CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

NARCIS (Netherlands)

Gessner, R.; Aiello, G.; Grossetti, G.; Meier, A.; Ronden, D.; Spaeh, P.; Scherer, T.; Schreck, S.; Strauss, D.; Vaccaro, A.

2013-01-01

The final design of the structural system for the ITER EC H&CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield

Conventional magnets. Pt. 1

International Nuclear Information System (INIS)

Marks, N.

1994-01-01

The design and construction of conventional, steel-cored, direct-current magnets are discussed. Laplace's equation and the associated cylindrical harmonic solutions in two dimensions are established. The equations are used to define the ideal pole shapes and required excitation for dipole, quadrupole and sextupole magnets. Standard magnet geometries are then considered and criteria determining the coil design are presented. The use of codes for predicting flux density distributions and the iterative techniques used for pole face design are then discussed. This includes a description of the use of two-dimensional codes to generate suitable magnet end geometries. Finally, standard constructional techniques for cores and coils are described. (orig.)

Updating the Design of the Poloidal Field Coils for the ITER Magnet System

International Nuclear Information System (INIS)

Yoshida, K.; Takahashi, Y.; Mitchell, N.; Jong, C.; Bessette, D.

2006-01-01

The ITER superconducting coil system consists of 18 Toroidal Field coils, six Poloidal Field (PF) coils, six Central Solenoid (CS) modules, 18 Correction Coils and their feeders. The six PF coils are attached to the TF coil cases through flexible plates or sliding supports allowing radial displacements. The PF coils and CS modules provide suitable magnetic fields for plasma shaping and position control. The PF coils use NbTi superconductor, cooled by supercritical helium. This gives a substantial cost saving compared to Nb 3 Sn and the elimination of a reaction heat treatment greatly simplifies the insulation of such large diameter coils. The cable configuration is 6 sub-cables arranged around a central cooling space. The conductors have a heavy square walled stainless steel jacket. The latest parameters of conductor design are evaluated by analysis of the minimum quench energy and hotspot temperature. The PF coils are self supporting as regards the radial magnetic loads. The vertical loads on each PF coil are transmitted to the TF coil cases. Load transmission is through flexible plates for the PF2 to PF5 coils or sliding supports for the PF1 and PF6 coils with fibreslip bearing surfaces. The supports for the PF winding consist of a set of clamping plates and stud bolts. The shape of the clamping plates has been designed to minimize stresses in the winding pack insulation. Bolts are pre-tensioned to keep pressure between the winding pack and clamping plate. Because of the difficulties in replacing the PF coils, the most unreliable component (the coil insulation) is designed with extra redundancy. There are two insulation layers with a thin metal screen in between. By monitoring the voltage of the intermediate screen, it is possible to detect an incipient short, defined as a short in only one of the two insulation layers. Adjustment of the screen voltage level may allow the shot growth to the stopped once it is detected. Alternately the faulty double pancake must

ITER concept definition. V.2

International Nuclear Information System (INIS)

1989-01-01

Volume II of the two volumes describing the concept definition of the International Thermonuclear Experimental Reactor deals with the ITER concept in technical depth, and covers all areas of design of the ITER tokamak. Included are an assessment of the current database for design, scoping studies, rationale for concepts selection, performance flexibility, the ITER concept, the operations and experimental/testing program, ITER parameters and design phase schedule, and research and development specific to ITER. This latter includes a definition of specific research and development tasks, a division of tasks among members, specific milestones, required results, and schedules. Figs and tabs

ITER CTA newsletter. No. 10

International Nuclear Information System (INIS)

2002-07-01

This ITER CTA newsletter issue comprises the ITER backgrounder, which was approved as an official document by the participants in the Negotiations on the ITER Implementation agreement at their fourth meeting, held in Cadarache from 4-6 June 2002, and information about two ITER meetings: one is the third meeting of the ITER parties' designated Safety Representatives, which took place in Cadarache, France from 6-7 June 2002, and the other is the second meeting of the International Tokamak Physics Activity (ITPA) topical group on diagnostics, which was held at General Atomics, San Diego, USA, from 4-8 March 2002

Conceptual design of fusion experimental reactor (FER/ITER)

International Nuclear Information System (INIS)

Kimura, Haruyuki; Saigusa, Mikio; Saitoh, Yasushi

1991-06-01

Conceptual design of the Ion Cyclotron Wave (ICW) system for FER and Japanese contribution to the conceptual design of the ITER ICW system are presented. A frequency range of the FER ICW system is 50-85 MHz, which covers 2ω cT heating, current drive by transit time magnetic pumping (TTMP) and 2ω cD heating. Physics analyses show that the FER and the ITER ICW systems are suitable for the central ion heating and the burn control. The launching systems of the FER ICW system and the ITER high frequency ICW system are characterized by in-port plug and ridged-waveguide-fed 5x4 phased loop array. Merits of those systems are (1) a ceramic support is not necessary inside the cryostat and (2) remote maintenance of the front end part of the launcher is relatively easy. Overall structure of the launching system is consistent with radiation shielding, cooling, pumping, tritium safety and remote maintenance. The launcher has injection capability of 20 MW in the frequency range of 50-85 MHz with the separatrix-antenna distance of 15 cm and steep scrape-off density profile of H-mode. The shape of the ridged waveguide is optimized to provide desired frequency range and power handling capability with a finite element method. Matching between the current strap and the ridged waveguide is satisfactorily good. Thermal analysis of the Faraday shield shows that high electric conductivity low Z material such as beryllium should be chosen for a protection tile of the Faraday shield. Thick Faraday shield is necessary to tolerate electromagnetic force during disruptions. R and D needs for the ITER/FER ICW systems are identified and gain from JT-60/60U ICRF experiments and operations are indicated in connection with them. (author)

Toward construction of ITER

International Nuclear Information System (INIS)

Shimomura, Yasuo

2005-01-01

The ITER Project has been significantly developed in the past years in preparation for its construction. The ITER Negotiators have developed a draft Joint Implementation Agreement (JIA), ready for completion following the nomination of the Project's Director General (DG). The ITER International Team and Participant Teams have continued technical and organizational preparations. The actual construction will be able to start immediately after the international ITER organization will be established, following signature of the JIA. The Project is now strongly supported by all the participants as well as by the scientific community with the final high-level negotiations, focused on siting and the concluding details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. The extent to which they both wish to host the ITER facility is such that large contributions to a broader collaboration among the Parties are also proposed by them. This covers complementary activities to help accelerate fusion development towards a viable power source, as well as may allow the Participants to reach a conclusion on ITER siting. (author)

Drifts, currents, and power scrape-off width in SOLPS-ITER modeling of DIII-D

International Nuclear Information System (INIS)

Meier, E. T.; Goldston, R. J.; Kaveeva, E. G.; Makowski, M. A.; Mordijck, S.

2016-01-01

The effects of drifts and associated flows and currents on the width of the parallel heat flux channel (λ q ) in the tokamak scrape-off layer (SOL) are analyzed using the SOLPS-ITER 2D fluid transport code. Motivation is supplied by Goldston’s heuristic drift (HD) model for λ q , which yields the same approximately inverse poloidal magnetic field dependence seen in multi-machine regression. The analysis, focusing on a DIII-D H-mode discharge, reveals HD-like features, including comparable density and temperature fall-off lengths in the SOL, and up-down ion pressure asymmetry that allows net cross-separatrix ion magnetic drift flux to exceed net anomalous ion flux. In experimentally relevant high-recycling cases, scans of both toroidal and poloidal magnetic field (B tor and B pol ) are conducted, showing minimal λ q dependence on either component of the field. Insensitivity to B tor is expected, and suggests that SOLPS-ITER is effectively capturing some aspects of HD physics. Absence of λ q dependence on B pol , however, is inconsistent with both the HD model and experimental results. As a result, the inconsistency is attributed to strong variation in the parallel Mach number, which violates one of the premises of the HD model.

Local control unit for ITER-India gyrotron test facility (IIGTF)

Energy Technology Data Exchange (ETDEWEB)

Rathod, Vipal, E-mail: vipal.rathod@iter-india.org; Shah, Ronak; Mandge, Deepak; Parmar, Rajvi; Rao, S.L.

2016-11-15

Highlights: • A dedicated full scale ITER prototype Local Control Unit for ITER-India Gyrotron test facility. • National Instruments® make PXIe system for real time control & data acquisition and Siemens® PLC for sequence control function. • Hardwired FPGA based fast protection interlock system. • High speed analog fiber optical transmission link using V/F and F/V technique. • Software framework based on LabVIEW™ platform and ITER CODAC Core System. - Abstract: Electron Cyclotron system on ITER, is one of the important RF ancillary systems based on high power Gyrotron RF sources, that is used for plasma heating and current drive applications. To operate a Gyrotron source, various auxiliary systems and services such as Super Conducting Magnet set, High Voltage Power Supplies, Auxiliary Power Supplies, Waveguide components, Cooling water system and a Local Control Unit (LCU) are required. The LCU plays a very crucial role for the safe and reliable operation of Gyrotron system. A dedicated full scale ITER prototype LCU is being developed for testing and commissioning of an ITER like Test Gyrotron at ITER-India Gyrotron Test facility (IIGTF). The main functions of LCU include Sequence Control, Local Interlock Protection and Real Time Data Acquisition. PLC based slow controller is used for implementing the Sequence Control & Slow Interlock functions. Critical Protection Interlocks are required to have a response time of <10 μs and are implemented using custom built hardware and PXIe based fast controller. Also PXIe system is used for implementing Real Time Data Acquisition function that is required to have slow and fast acquisition with online visualization and off line analysis facility. A Signal Conditioning Unit (SCU) is used to interface and faithfully transmit the field signals to the remote control systems. Necessary controller hardware is procured and several pre-prototype developments have been taken up to establish the critical subsystems such as

Local control unit for ITER-India gyrotron test facility (IIGTF)

International Nuclear Information System (INIS)

Rathod, Vipal; Shah, Ronak; Mandge, Deepak; Parmar, Rajvi; Rao, S.L.

2016-01-01

Highlights: • A dedicated full scale ITER prototype Local Control Unit for ITER-India Gyrotron test facility. • National Instruments® make PXIe system for real time control & data acquisition and Siemens® PLC for sequence control function. • Hardwired FPGA based fast protection interlock system. • High speed analog fiber optical transmission link using V/F and F/V technique. • Software framework based on LabVIEW™ platform and ITER CODAC Core System. - Abstract: Electron Cyclotron system on ITER, is one of the important RF ancillary systems based on high power Gyrotron RF sources, that is used for plasma heating and current drive applications. To operate a Gyrotron source, various auxiliary systems and services such as Super Conducting Magnet set, High Voltage Power Supplies, Auxiliary Power Supplies, Waveguide components, Cooling water system and a Local Control Unit (LCU) are required. The LCU plays a very crucial role for the safe and reliable operation of Gyrotron system. A dedicated full scale ITER prototype LCU is being developed for testing and commissioning of an ITER like Test Gyrotron at ITER-India Gyrotron Test facility (IIGTF). The main functions of LCU include Sequence Control, Local Interlock Protection and Real Time Data Acquisition. PLC based slow controller is used for implementing the Sequence Control & Slow Interlock functions. Critical Protection Interlocks are required to have a response time of <10 μs and are implemented using custom built hardware and PXIe based fast controller. Also PXIe system is used for implementing Real Time Data Acquisition function that is required to have slow and fast acquisition with online visualization and off line analysis facility. A Signal Conditioning Unit (SCU) is used to interface and faithfully transmit the field signals to the remote control systems. Necessary controller hardware is procured and several pre-prototype developments have been taken up to establish the critical subsystems such as

IVVS actuating system compatibility test to ITER gamma radiation conditions

Energy Technology Data Exchange (ETDEWEB)

Rossi, Paolo, E-mail: paolo.rossi@enea.it [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Collibus, M. Ferri de; Florean, M.; Monti, C.; Mugnaini, G.; Neri, C.; Pillon, M.; Pollastrone, F. [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Baccaro, S.; Piegari, A. [ENEA CR Casaccia, 301 Via Anguillarese, 00123 Santa Maria di Galeria, Rome (Italy); Damiani, C.; Dubus, G. [Fusion For Energy c/Josep Pla, n° 2 Torres Diagonal Litoral, 08019 Barcelona (Spain)

2013-10-15

Highlights: • ENEA developed and tested a prototype of a laser In Vessel Viewing and ranging System (IVVS) for ITER. • One piezo-motor prototype has been tested on the ENEA Calliope gamma irradiation facility to verify its compatibility to ITER gamma radiation conditions. • After a total dose of more than 4 MGy the piezo-motor maintained almost the same working parameters monitored before test without any evident and significant degradation of functionality. • After the full gamma irradiation test, the same piezo-motor assembly will be tested with 14 MeV neutrons irradiation using ENEA FNG facility. -- Abstract: The In Vessel Viewing System (IVVS) is a fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A design and testing activity is ongoing, in the framework of a Fusion for Energy (F4E) grant agreement, to make the IVVS probe design compatible with ITER operating conditions and in particular, but not only, with attention to neutrons and gammas fluxes and both space constraints and interfaces. The paper describes the testing activity performed on the customized piezoelectric motors and the main components of the actuating system of the IVVS probe with reference to ITER gamma radiation conditions. In particular the test is performed on the piezoelectric motor, optical encoder and small scale optical samples .The test is carried out on the ENEA Calliope gamma irradiation facility at ITER relevant gamma fields at rate of about 2.5 kGy/h and doses of 4 MGy. The paper reports in detail the setup arrangement of the test campaign in order to verify significant working capability of the IVVS actuating components and the results are shown in terms of functional performances and parameters. The overall test campaign on IVVS actuating system will be completed on other ENEA testing facilities in order to verify compatibility to Magnetic field, neutrons and thermal

IVVS actuating system compatibility test to ITER gamma radiation conditions

International Nuclear Information System (INIS)

Rossi, Paolo; Collibus, M. Ferri de; Florean, M.; Monti, C.; Mugnaini, G.; Neri, C.; Pillon, M.; Pollastrone, F.; Baccaro, S.; Piegari, A.; Damiani, C.; Dubus, G.

2013-01-01

Highlights: • ENEA developed and tested a prototype of a laser In Vessel Viewing and ranging System (IVVS) for ITER. • One piezo-motor prototype has been tested on the ENEA Calliope gamma irradiation facility to verify its compatibility to ITER gamma radiation conditions. • After a total dose of more than 4 MGy the piezo-motor maintained almost the same working parameters monitored before test without any evident and significant degradation of functionality. • After the full gamma irradiation test, the same piezo-motor assembly will be tested with 14 MeV neutrons irradiation using ENEA FNG facility. -- Abstract: The In Vessel Viewing System (IVVS) is a fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A design and testing activity is ongoing, in the framework of a Fusion for Energy (F4E) grant agreement, to make the IVVS probe design compatible with ITER operating conditions and in particular, but not only, with attention to neutrons and gammas fluxes and both space constraints and interfaces. The paper describes the testing activity performed on the customized piezoelectric motors and the main components of the actuating system of the IVVS probe with reference to ITER gamma radiation conditions. In particular the test is performed on the piezoelectric motor, optical encoder and small scale optical samples .The test is carried out on the ENEA Calliope gamma irradiation facility at ITER relevant gamma fields at rate of about 2.5 kGy/h and doses of 4 MGy. The paper reports in detail the setup arrangement of the test campaign in order to verify significant working capability of the IVVS actuating components and the results are shown in terms of functional performances and parameters. The overall test campaign on IVVS actuating system will be completed on other ENEA testing facilities in order to verify compatibility to Magnetic field, neutrons and thermal

Benchmarking of MCAM 4.0 with the ITER 3D Model

International Nuclear Information System (INIS)

Ying Li; Lei Lu; Aiping Ding; Haimin Hu; Qin Zeng; Shanliang Zheng; Yican Wu

2006-01-01

Monte Carlo particle transport simulations are widely employed in fields such as nuclear engineering, radio-therapy and space science. Describing and verifying the 3D geometry of fusion devices, however, are among the most complex tasks of MCNP calculation problems in nuclear analysis. The manual modeling of a complex geometry for MCNP code, though a common practice, is an extensive, time-consuming, and error prone task. An efficient solution is to shift the geometric modeling into Computer Aided Design(CAD) systems and to use an interface for MCNP to convert the CAD model to MCNP file. The advantage of this approach lies in the fact that it allows access to full features of modern CAD systems facilitating the geometric modeling and utilizing the existing CAD models. MCAM(MCNP Automatic Modeling System) is an integrated tool for CAD model preprocessing, accurate bi-directional conversion between CAD/MCNP models, neutronics property processing and geometric modeling developed by FDS team in ASIPP and Hefei University of Technology. MCAM4.0 has been extended and enhanced to support various CAD file formats and the preprocessing of CAD model, such as healing, automatic model reconstruction, overlap detection and correction, automatic void modeling. The ITER international benchmark model is provided by ITER international team to compare the CAD/MCNP programs being developed in the ITER participant teams. It is created in CATIA/V5, which has been chosen as the CAD system for ITER design, including all the important parts and components of the ITER device. The benchmark model contains vast curve surfaces, which can fully test the ability of MCNP/CAD codes. The whole processing procedure of this model will be presented in this paper, which includes the geometric model processing, neutroics property processing, converting to MCNP input file, calculating with MCNP and analysis. The nuclear analysis results of the model will be given in the end. Although these preliminary

ITER task T26/28 (1994): preliminary results on the solubility, diffusion and permeability of hydrogen isotopes in potentional fusion reactor ceramics

International Nuclear Information System (INIS)

Thompson, D.; Macauley-Newcombe, R.

1995-02-01

Ceramic insulators are integral parts of numerous components essential for the heating, control and diagnostic measurement of fusion plasmas. For safe and reliable reactor operations it is important to be able to predict the resultant tritium inventories and permeation fluxes. Currently there is little or no published data on tritium behaviour in AlN. This report contains the preliminary results of work begun in 1994 on ITER task T26/28. The ceramics studied in 1994 were AlN and Al 2 O 3 . Section 2.1 reports on the measurements of permeation through an AlN film on a vanadium substrate, with supplementary Rutherford backscattering measurements of the surface composition, before and after permeation. Section 2.2 deals with ion-beam analysis of hydrogen and deuterium in sapphire and alumina, before and after room temperature implantation of deuterium, with careful attention to the effects of the analysis ions upon the data; section 2.3 looks at the data in 2.2 from the perspective of ion-beam-induced desorption; and in section 2.4 a thermal desorption measurement is reported for comparison. The numbers derived include effective diffusivities and, in the case of AlN, an estimated solubility, for hydrogen isotopes. 16 refs., 1 tab., 12 figs

Calculation of magnetic field and electromagnetic forces in MHD superconducting magnets

International Nuclear Information System (INIS)

Martinelli, G.; Morini, A.; Moisio, M.F.

1992-01-01

The realization of a superconducting prototype magnet for MHD energy conversion is under development in Italy. Electromechanical industries and University research groups are involved in the project. The paper deals with analytical methods developed at the Department of Electrical Engineering of Padova University for calculating magnetic field and electromagnetic forces in MHD superconducting magnets and utilized in the preliminary design of the prototype

Plasma edge control by chaotic magnetic field structures. Book of abstracts

International Nuclear Information System (INIS)

2013-01-01

The following topics were dealt with: Formation of stochastic magnetic layers and plasma response to external, non-axisymmetric magnetic perturbations, energy and particle transport in stochastic magnetic fields and 3D equilibria, application of resonant magnetic perturbations for ELM control and implications for ITER, transport and exhaust in helical and island divertors. (HSI)

Compressively sampled MR image reconstruction using generalized thresholding iterative algorithm

Science.gov (United States)

Elahi, Sana; kaleem, Muhammad; Omer, Hammad

2018-01-01

Compressed sensing (CS) is an emerging area of interest in Magnetic Resonance Imaging (MRI). CS is used for the reconstruction of the images from a very limited number of samples in k-space. This significantly reduces the MRI data acquisition time. One important requirement for signal recovery in CS is the use of an appropriate non-linear reconstruction algorithm. It is a challenging task to choose a reconstruction algorithm that would accurately reconstruct the MR images from the under-sampled k-space data. Various algorithms have been used to solve the system of non-linear equations for better image quality and reconstruction speed in CS. In the recent past, iterative soft thresholding algorithm (ISTA) has been introduced in CS-MRI. This algorithm directly cancels the incoherent artifacts produced because of the undersampling in k -space. This paper introduces an improved iterative algorithm based on p -thresholding technique for CS-MRI image reconstruction. The use of p -thresholding function promotes sparsity in the image which is a key factor for CS based image reconstruction. The p -thresholding based iterative algorithm is a modification of ISTA, and minimizes non-convex functions. It has been shown that the proposed p -thresholding iterative algorithm can be used effectively to recover fully sampled image from the under-sampled data in MRI. The performance of the proposed method is verified using simulated and actual MRI data taken at St. Mary's Hospital, London. The quality of the reconstructed images is measured in terms of peak signal-to-noise ratio (PSNR), artifact power (AP), and structural similarity index measure (SSIM). The proposed approach shows improved performance when compared to other iterative algorithms based on log thresholding, soft thresholding and hard thresholding techniques at different reduction factors.

Upper limb stroke rehabilitation: the effectiveness of Stimulation Assistance through Iterative Learning (SAIL).

Science.gov (United States)

Meadmore, Katie L; Cai, Zhonglun; Tong, Daisy; Hughes, Ann-Marie; Freeman, Chris T; Rogers, Eric; Burridge, Jane H

2011-01-01

A novel system has been developed which combines robotic therapy with electrical stimulation (ES) for upper limb stroke rehabilitation. This technology, termed SAIL: Stimulation Assistance through Iterative Learning, employs advanced model-based iterative learning control (ILC) algorithms to precisely assist participant's completion of 3D tracking tasks with their impaired arm. Data is reported from a preliminary study with unimpaired participants, and also from a single hemiparetic stroke participant with reduced upper limb function who has used the system in a clinical trial. All participants completed tasks which involved moving their (impaired) arm to follow an image of a slowing moving sphere along a trajectory. The participants' arm was supported by a robot and ES was applied to the triceps brachii and anterior deltoid muscles. During each task, the same tracking trajectory was repeated 6 times and ILC was used to compute the stimulation signals to be applied on the next iteration. Unimpaired participants took part in a single, one hour training session and the stroke participant undertook 18, 1 hour treatment sessions composed of tracking tasks varying in length, orientation and speed. The results reported describe changes in tracking ability and demonstrate feasibility of the SAIL system for upper limb rehabilitation. © 2011 IEEE

Estimation of error fields from ferromagnetic parts in ITER

Energy Technology Data Exchange (ETDEWEB)

Oliva, A. Bonito [Fusion for Energy (Spain); Chiariello, A.G.; Formisano, A.; Martone, R. [Ass. EURATOM/ENEA/CREATE, Dip. di Ing. Industriale e dell’Informazione, Seconda Università di Napoli, Via Roma 29, I-81031 Napoli (Italy); Portone, A., E-mail: alfredo.portone@f4e.europa.eu [Fusion for Energy (Spain); Testoni, P. [Fusion for Energy (Spain)

2013-10-15

Highlights: ► The paper deals with error fields generated in ITER by magnetic masses. ► Magnetization state is computed from simplified FEM models. ► Closed form expressions adopted for the flux density of magnetized parts are given. ► Such expressions allow to simplify the estimation of the effect of iron pieces (or lack of) on error field. -- Abstract: Error fields in tokamaks are small departures from the exact axisymmetry of the ideal magnetic field configuration. Their reduction below a threshold value by the error field correction coils is essential since sufficiently large static error fields lead to discharge disruption. The error fields are originated not only by magnets fabrication and installation tolerances, by the joints and by the busbars, but also by the presence of ferromagnetic elements. It was shown that superconducting joints, feeders and busbars play a secondary effect; however in order to estimate of the importance of each possible error field source, rough evaluations can be very useful because it can provide an order of magnitude of the correspondent effect and, therefore, a ranking in the request for in depth analysis. The paper proposes a two steps procedure. The first step aims to get the approximate magnetization state of ferromagnetic parts; the second aims to estimate the full 3D error field over the whole volume using equivalent sources for magnetic masses and taking advantage from well assessed approximate closed form expressions, well suited for the far distance effects.

Structural analyses of ITER toroidal field coils under fault conditions

International Nuclear Information System (INIS)

Jong, C.T.J.

1992-04-01

ITER (International Thermonuclear Experimental Reactor) is intended to be an experimental thermonuclear tokamak reactor testing the basic physics performance and technologies essential to future fusion reactors. The magnet system of ITER consists essentially of 4 sub-systems, i.e. toroidal field coils (TFCs), poloidal field coils (PFCs), power supplies, and cryogenic supplies. These subsystems do not contain significant radioactivity inventories, but the large energy inventory is a potential accident initiator. The aim of the structural analyses is to prevent accidents from propagating into vacuum vessel, tritium system and cooling system, which all contain significant amounts of radioactivity. As part of design process 3 conditions are defined for PF and TF coils, at which mechanical behaviour has to be analyzed in some detail, viz: normal operating conditions, upset conditions and fault conditions. This paper describes the work carried out by ECN to create a detailed finite element model of 16 TFCs as well as results of some fault condition analyses made with the model. Due to fault conditions, either electrical or mechanical, magnetic loading of TFCs becomes abnormal and further mechanical failure of parts of the overall structure might occur (e.g. failure of coil, gravitational supports, intercoil structure). The analyses performed consist of linear elastic stress analyses and electro-magneto-structural analyses (coupled field analyses). 8 refs.; 5 figs.; 5 tabs

ITER technology R and D during the EDA

International Nuclear Information System (INIS)

Mizoguchi, T.

2001-01-01

A short overview of the ITER technology R and D achievements is presented. It includes R and D programme in the area of superconducting magnets, L-1 central solenoid model coil, L-2 toroidal field model coil, L-3 vacuum vessel sector, L-4 blanket module, L-5 divertor cassette, L-6 blanket and L-7 divertor remote handling systems. In addition to the seven large R and D projects, development of components for fuelling, pumping, tritium processing, heating/current drive, power supplies and plasma diagnostics, as well as safety-related R and D have significantly progressed

Validation of the model for ELM suppression with 3D magnetic fields using low torque ITER baseline scenario discharges in DIII-D

Science.gov (United States)

Moyer, R. A.; Paz-Soldan, C.; Nazikian, R.; Orlov, D. M.; Ferraro, N. M.; Grierson, B. A.; Knölker, M.; Lyons, B. C.; McKee, G. R.; Osborne, T. H.; Rhodes, T. L.; Meneghini, O.; Smith, S.; Evans, T. E.; Fenstermacher, M. E.; Groebner, R. J.; Hanson, J. M.; La Haye, R. J.; Luce, T. C.; Mordijck, S.; Solomon, W. M.; Turco, F.; Yan, Z.; Zeng, L.; DIII-D Team

2017-10-01

Experiments have been executed in the DIII-D tokamak to extend suppression of Edge Localized Modes (ELMs) with Resonant Magnetic Perturbations (RMPs) to ITER-relevant levels of beam torque. The results support the hypothesis for RMP ELM suppression based on transition from an ideal screened response to a tearing response at a resonant surface that prevents expansion of the pedestal to an unstable width [Snyder et al., Nucl. Fusion 51, 103016 (2011) and Wade et al., Nucl. Fusion 55, 023002 (2015)]. In ITER baseline plasmas with I/aB = 1.4 and pedestal ν * ˜ 0.15, ELMs are readily suppressed with co- I p neutral beam injection. However, reducing the beam torque from 5 Nm to ≤ 3.5 Nm results in loss of ELM suppression and a shift in the zero-crossing of the electron perpendicular rotation ω ⊥ e ˜ 0 deeper into the plasma. The change in radius of ω ⊥ e ˜ 0 is due primarily to changes to the electron diamagnetic rotation frequency ωe * . Linear plasma response modeling with the resistive MHD code m3d-c1 indicates that the tearing response location tracks the inward shift in ω ⊥ e ˜ 0. At pedestal ν * ˜ 1, ELM suppression is also lost when the beam torque is reduced, but the ω ⊥ e change is dominated by collapse of the toroidal rotation v T . The hypothesis predicts that it should be possible to obtain ELM suppression at reduced beam torque by also reducing the height and width of the ωe * profile. This prediction has been confirmed experimentally with RMP ELM suppression at 0 Nm of beam torque and plasma normalized pressure β N ˜ 0.7. This opens the possibility of accessing ELM suppression in low torque ITER baseline plasmas by establishing suppression at low beta and then increasing beta while relying on the strong RMP-island coupling to maintain suppression.

Power converters for ITER

CERN Document Server

Benfatto, I

2006-01-01

The International Thermonuclear Experimental Reactor (ITER) is a thermonuclear fusion experiment designed to provide long deuterium– tritium burning plasma operation. After a short description of ITER objectives, the main design parameters and the construction schedule, the paper describes the electrical characteristics of the French 400 kV grid at Cadarache: the European site proposed for ITER. Moreover, the paper describes the main requirements and features of the power converters designed for the ITER coil and additional heating power supplies, characterized by a total installed power of about 1.8 GVA, modular design with basic units up to 90 MVA continuous duty, dc currents up to 68 kA, and voltages from 1 kV to 1 MV dc.

In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

Energy Technology Data Exchange (ETDEWEB)

L. C. Cadwallader

2013-01-01

The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

Mechanical Properties of High Manganese Austenitic Stainless Steel JK2LB for ITER Central Solenoid Jacket Material

Science.gov (United States)

Saito, Toru; Kawano, Katsumi; Yamazaki, Toru; Ozeki, Hidemasa; Isono, Takaaki; Hamada, Kazuya; Devred, Arnaud; Vostner, Alexander

A suite of advanced austenitic stainless steels are used for the ITER TF, CS and PF coil systems.These materials will be exposed to cyclic-stress at cryogenic temperature. Therefore, high manganese austenitic stainless steel JK2LB, which has high tensile strength, high ductility and high resistance to fatigue at 4 K has been chosen for the CS conductor. The cryogenic temperature mechanical property data of this material are very important for the ITER magnet design. This study is focused on mechanical characteristics of JK2LB and its weld joint.

ITER vacuum vessel design (D201 subtask 1.3 and subtask 3). Final report

International Nuclear Information System (INIS)

1996-01-01

ITER Task No. D201, Vacuum Vessel Design (Subtask 1.3 and Subtask 3), was initiated to propose and evaluate local vacuum vessel reinforcement alternatives in proximity to the Neutral Beam, Radial Mid-Plane, Top, and Divertor Ports. These areas were reported to be highly stressed regions based on the results of preliminary stress analyses performed by the USHT (US Home Team) and the ITER Joint Central Team (JCT) at the Garching JWS (Joint Work Site). Initial design activities focused on the divertor port region which was reported to experience the highest stress intensities. Existing stress analysis models and results were reviewed with the USHT stress analysts to obtain an overall understanding of the vessel response to the various applied loads. These reviews indicated that the reported stress intensities in the divertor port region were significantly affected by the loads applied to the vessel in adjacent regions

Impact of knowledge-based iterative model reconstruction on myocardial late iodine enhancement in computed tomography and comparison with cardiac magnetic resonance.

Science.gov (United States)

Tanabe, Yuki; Kido, Teruhito; Kurata, Akira; Fukuyama, Naoki; Yokoi, Takahiro; Kido, Tomoyuki; Uetani, Teruyoshi; Vembar, Mani; Dhanantwari, Amar; Tokuyasu, Shinichi; Yamashita, Natsumi; Mochizuki, Teruhito

2017-10-01

We evaluated the image quality and diagnostic performance of late iodine enhancement computed tomography (LIE-CT) with knowledge-based iterative model reconstruction (IMR) for the detection of myocardial infarction (MI) in comparison with late gadolinium enhancement magnetic resonance imaging (LGE-MRI). The study investigated 35 patients who underwent a comprehensive cardiac CT protocol and LGE-MRI for the assessment of coronary artery disease. The CT protocol consisted of stress dynamic myocardial CT perfusion, coronary CT angiography (CTA) and LIE-CT using 256-slice CT. LIE-CT scans were acquired 5 min after CTA without additional contrast medium and reconstructed with filtered back projection (FBP), a hybrid iterative reconstruction (HIR), and IMR. The signal-to-noise ratio (SNR) and contrast-to-noise ratio (CNR) were assessed. Sensitivity and specificity of LIE-CT for detecting MI were assessed according to the 16-segment model. Image quality scores, and diagnostic performance were compared among LIE-CT with FBP, HIR and IMR. Among the 35 patients, 139 of 560 segments showed MI in LGE-MRI. On LIE-CT with FBP, HIR, and IMR, the median SNRs were 2.1, 2.9, and 6.1; and the median CNRs were 1.7, 2.2, and 4.7, respectively. Sensitivity and specificity were 56 and 93% for FBP, 62 and 91% for HIR, and 80 and 91% for IMR. LIE-CT with IMR showed the highest image quality and sensitivity (p quality and diagnostic performance of LIE-CT for detecting MI in comparison with FBP and HIR.

Magnetic field measurements of the superEBIS superconducting magnet

International Nuclear Information System (INIS)

Herschcovitch, A.; Kponou, A.; Clipperton, R.; Hensel, W.; Usack, F.

1994-01-01

SuperEBIS was designed to have a solenoidal magnetic field of a 5 Tesla strength with a 120 cm long bore. The field was specified to be straight within 1 part in 10000 within the bore, and uniform to within 1 part in 1000 within the central 90 cm. Magnetic field measurements were performed with a computerized magnetic field measuring setup that was borrowed from W. Sampson's group. A preliminary test was made of a scheme to determine if the magnetic and mechanical axes of the solenoid coincided, and, if not, by how much

Project management techniques used in the European Vacuum Vessel sectors procurement for ITER

Energy Technology Data Exchange (ETDEWEB)

Losasso, Marcello, E-mail: marcello.losasso@f4e.europa.eu [Fusion for Energy (F4E), Barcelona (Spain); Ortiz de Zuniga, Maria; Jones, Lawrence; Bayon, Angel; Arbogast, Jean-Francois; Caixas, Joan; Fernandez, Jose; Galvan, Stefano; Jover, Teresa [Fusion for Energy (F4E), Barcelona (Spain); Ioki, Kimihiro [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lewczanin, Michal; Mico, Gonzalo; Pacheco, Jose Miguel [Fusion for Energy (F4E), Barcelona (Spain); Preble, Joseph [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Stamos, Vassilis; Trentea, Alexandru [Fusion for Energy (F4E), Barcelona (Spain)

2012-08-15

Highlights: Black-Right-Pointing-Pointer File name contains the directory tree structure with a string of three-letter acronyms, thereby enabling parent directory location when confronted with orphan files. Black-Right-Pointing-Pointer The management of the procurement procedure was carried out in an efficient and timely manner, achieving precisely the contract placement date foreseen at the start of the process. Black-Right-Pointing-Pointer The contract start-up has been effectively implemented and a flexible project management system has been put in place for an efficient monitoring of the contract. - Abstract: The contract for the seven European Sectors of the ITER Vacuum Vessel (VV) was placed at the end of 2010 with a consortium of three Italian companies. The task of placing and the initial take-off of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. A stringent quality controlled system with a bespoke Vacuum Vessel Project Lifecycle Management system to control the information flow, based on ENOVIA SmarTeam, was developed to handle the storage and approval of Documentation including links to the F4E Vacuum Vessel system and ITER International Organization System interfaces. The VV Sector design and manufacturing schedule is based on Primavera software, which is cost loaded thus allowing F4E to carry out performance measurement with respect to its payments and commitments. This schedule is then integrated into the overall Vacuum Vessel schedule, which includes ancillary activities such as instruments, preliminary design and analysis. The VV Sector Risk Management included three separate risk analyses from F4E and the bidders, utilizing two different methodologies. These efforts will lead to an efficient and effective implementation of this contract, vital to the success of the ITER machine, since the Vacuum Vessel is the biggest single work package of Europe's contribution to ITER and

Iteration and accelerator dynamics

International Nuclear Information System (INIS)

Peggs, S.

1987-10-01

Four examples of iteration in accelerator dynamics are studied in this paper. The first three show how iterations of the simplest maps reproduce most of the significant nonlinear behavior in real accelerators. Each of these examples can be easily reproduced by the reader, at the minimal cost of writing only 20 or 40 lines of code. The fourth example outlines a general way to iteratively solve nonlinear difference equations, analytically or numerically

An Iterative Interplanetary Scintillation (IPS) Analysis Using Time-dependent 3-D MHD Models as Kernels

Science.gov (United States)

Jackson, B. V.; Yu, H. S.; Hick, P. P.; Buffington, A.; Odstrcil, D.; Kim, T. K.; Pogorelov, N. V.; Tokumaru, M.; Bisi, M. M.; Kim, J.; Yun, J.

2017-12-01

The University of California, San Diego has developed an iterative remote-sensing time-dependent three-dimensional (3-D) reconstruction technique which provides volumetric maps of density, velocity, and magnetic field. We have applied this technique in near real time for over 15 years with a kinematic model approximation to fit data from ground-based interplanetary scintillation (IPS) observations. Our modeling concept extends volumetric data from an inner boundary placed above the Alfvén surface out to the inner heliosphere. We now use this technique to drive 3-D MHD models at their inner boundary and generate output 3-D data files that are fit to remotely-sensed observations (in this case IPS observations), and iterated. These analyses are also iteratively fit to in-situ spacecraft measurements near Earth. To facilitate this process, we have developed a traceback from input 3-D MHD volumes to yield an updated boundary in density, temperature, and velocity, which also includes magnetic-field components. Here we will show examples of this analysis using the ENLIL 3D-MHD and the University of Alabama Multi-Scale Fluid-Kinetic Simulation Suite (MS-FLUKSS) heliospheric codes. These examples help refine poorly-known 3-D MHD variables (i.e., density, temperature), and parameters (gamma) by fitting heliospheric remotely-sensed data between the region near the solar surface and in-situ measurements near Earth.

IHadoop: Asynchronous iterations for MapReduce

KAUST Repository

Elnikety, Eslam Mohamed Ibrahim

2011-11-01

MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop\\'s task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application\\'s latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

IHadoop: Asynchronous iterations for MapReduce

KAUST Repository

Elnikety, Eslam Mohamed Ibrahim; El Sayed, Tamer S.; Ramadan, Hany E.

2011-01-01

MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop's task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application's latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

ITER ITA newsletter No. 31, June 2006

International Nuclear Information System (INIS)

2006-07-01

This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about initialling the ITER Agreement and its related instruments by seven ITER parties, which too place in Brussels on 24 May 2006. The initialling constituted the final act of the ITER negotiations. It confirmed the Parties' common acceptance of the negotiated texts, ad referendum, and signalled their intentions to move forward towards the entry into force of the ITER Agreement as soon as possible. 'ITER - Uniting science today, global energy tomorrow' was the theme of a number of media events timed to accompany a remarkable day in the history of the ITER international venture, May 24th 2006, initialling of the ITER international agreement

Enhanced nonlinear iterative techniques applied to a nonequilibrium plasma flow

International Nuclear Information System (INIS)

Knoll, D.A.

1998-01-01

The authors study the application of enhanced nonlinear iterative methods to the steady-state solution of a system of two-dimensional convection-diffusion-reaction partial differential equations that describe the partially ionized plasma flow in the boundary layer of a tokamak fusion reactor. This system of equations is characterized by multiple time and spatial scales and contains highly anisotropic transport coefficients due to a strong imposed magnetic field. They use Newton's method to linearize the nonlinear system of equations resulting from an implicit, finite volume discretization of the governing partial differential equations, on a staggered Cartesian mesh. The resulting linear systems are neither symmetric nor positive definite, and are poorly conditioned. Preconditioned Krylov iterative techniques are employed to solve these linear systems. They investigate both a modified and a matrix-free Newton-Krylov implementation, with the goal of reducing CPU cost associated with the numerical formation of the Jacobian. A combination of a damped iteration, mesh sequencing, and a pseudotransient continuation technique is used to enhance global nonlinear convergence and CPU efficiency. GMRES is employed as the Krylov method with incomplete lower-upper (ILU) factorization preconditioning. The goal is to construct a combination of nonlinear and linear iterative techniques for this complex physical problem that optimizes trade-offs between robustness, CPU time, memory requirements, and code complexity. It is shown that a mesh sequencing implementation provides significant CPU savings for fine grid calculations. Performance comparisons of modified Newton-Krylov and matrix-free Newton-Krylov algorithms will be presented

Status of the ITER EDA

International Nuclear Information System (INIS)

Aymar, R.

2000-01-01

This article summarizes progress made in the ITER Engineering Design Activities in the period between the ITER Meeting in Tokyo (January 2000) and June 2000. Topics: Termination of EDA, Joint Central Team and Support, Task Assignments, ITER Physics, Urgent and High Priority Physics Research Areas

Preliminary neutronic assessments for the development of the VIS/IR diagnostic systems located in the ITER EPP

International Nuclear Information System (INIS)

Palermo, Iole; Mota, Fernando; Rios, Luis; Catalán, Juan Pablo; Alonso, Javier; Ibarra, Angel

2015-01-01

Graphical abstract: - Highlights: • Neutronic and activation calculations for the VIS/IR ITER diagnostic. • Studied if silver could be used as a covering material for the Interspace components. • Determined the irradiation time in a gamma facility to test the vacuum window. • Neutron and gamma dose rate maps in the Port Area for proposed substrate and coatings. - Abstract: The paper focuses on the nuclear analyses of the ITER Equatorial Port Visible/Infrared Wide Angle Viewing System (VIS/IR WAVS). This instrument comprises of viewing systems in the 4 Equatorial Ports (EP) 3, 9, 12 and 17. The main mission of this diagnostic is to support the operation of the tokamak by providing visible and infrared viewing and temperature data of the first wall to protect it from damage. Its design is driven by both the tokamak severe environment and the high performances required for machine protection. New nuclear studies have been carried out for the development of the diagnostic and for test purposes under ITER-like irradiation conditions in order to choose the most appropriate materials for the optical components. Thus, three neutronic analyses have been carried out: the first in order to verify if silver could be used as a covering material for the optical components in different location of the Interspace area; the second in order to establish the irradiation time required in a Co-60 gamma facility (at CIEMAT) for testing purposes of the sapphire vacuum window; and the third to give more detailed specifications for the irradiation campaigns under gamma (in the Co-60 facility) and neutrons (at SCK·CEN BR2 reactor), about the time required to achieve the same dose than the one accumulated in ITER at the end-of-life (EOL) in the different components of the Port Area for the materials proposed as substrate and coatings. The neutronic and activation calculations have been performed using the Monte Carlo code MCNP5, the activation code ACAB and the cross section

Preliminary neutronic assessments for the development of the VIS/IR diagnostic systems located in the ITER EPP

Energy Technology Data Exchange (ETDEWEB)

Palermo, Iole, E-mail: iole.palermo@ciemat.es [CIEMAT, Fusion National Laboratory, Av. Complutense 40, E-28040 Madrid (Spain); Mota, Fernando; Rios, Luis [CIEMAT, Fusion National Laboratory, Av. Complutense 40, E-28040 Madrid (Spain); Catalán, Juan Pablo [UNED, Department of Energy Engineering, c/ Juan del Rosal 12, E-28040 Madrid (Spain); Alonso, Javier; Ibarra, Angel [CIEMAT, Fusion National Laboratory, Av. Complutense 40, E-28040 Madrid (Spain)

2015-11-15

Graphical abstract: - Highlights: • Neutronic and activation calculations for the VIS/IR ITER diagnostic. • Studied if silver could be used as a covering material for the Interspace components. • Determined the irradiation time in a gamma facility to test the vacuum window. • Neutron and gamma dose rate maps in the Port Area for proposed substrate and coatings. - Abstract: The paper focuses on the nuclear analyses of the ITER Equatorial Port Visible/Infrared Wide Angle Viewing System (VIS/IR WAVS). This instrument comprises of viewing systems in the 4 Equatorial Ports (EP) 3, 9, 12 and 17. The main mission of this diagnostic is to support the operation of the tokamak by providing visible and infrared viewing and temperature data of the first wall to protect it from damage. Its design is driven by both the tokamak severe environment and the high performances required for machine protection. New nuclear studies have been carried out for the development of the diagnostic and for test purposes under ITER-like irradiation conditions in order to choose the most appropriate materials for the optical components. Thus, three neutronic analyses have been carried out: the first in order to verify if silver could be used as a covering material for the optical components in different location of the Interspace area; the second in order to establish the irradiation time required in a Co-60 gamma facility (at CIEMAT) for testing purposes of the sapphire vacuum window; and the third to give more detailed specifications for the irradiation campaigns under gamma (in the Co-60 facility) and neutrons (at SCK·CEN BR2 reactor), about the time required to achieve the same dose than the one accumulated in ITER at the end-of-life (EOL) in the different components of the Port Area for the materials proposed as substrate and coatings. The neutronic and activation calculations have been performed using the Monte Carlo code MCNP5, the activation code ACAB and the cross section

ITER EDA newsletter. V. 10, special issue

International Nuclear Information System (INIS)

2001-07-01

This ITER EDA Newsletter includes summaries of the reports of ITER EDA JCT Physics unit about ITER physics R and D during the Engineering Design Activities (EDA), ITER EDA JCT Naka JWC ITER technology R and D during the EDA, and Safety, Environment and Health group of ITER EDA JCT, Garching JWS on EDA activities related to safety

Overview of ITER project. Objective, history, status of development, structure and its future

International Nuclear Information System (INIS)

Matsuda, S.

2003-01-01

The objective of ITER Project is to demonstrate the scientific and technological feasibility of fusion energy for peaceful use. To achieve this goal, the Conceptual Design Activities (CDA 1988-1990) and the Engineering Design Activities (EDA 1992-2001) were conducted by the joint effort of EU, Japan, RF and US under the auspices of IAEA. Engineering design has been developed consistently with plasma physics databases obtained by the world-wide magnetic fusion devices up to and during the EDA. Technology R and D including development, manufacturing and testing of scalable models of major components have been also conducted and ensured manufacturing and technical feasibility of ITER. Such technical readiness to start ITER construction urged formal government negotiation initially by Canada, EU, Japan and RF, and intense meetings have been held since November 2001. The US and China have joined since Feb. 2003. Four candidate sites have been proposed by Canada, EU and Japan for construction and have been evaluated. Once the agreement will be made, the construction takes about 10 years to complete, followed by the operation and exploitation phase lasting about 20 years. This paper gives an overview of ITER Project including its role in the long range fusion program, history, status of development, collaboration structure and its future program. (author)

Development of the ITER Continuous External Rogowski: From conceptual design to final design

Energy Technology Data Exchange (ETDEWEB)

Moreau, Philippe, E-mail: philippe.jacques.moreau@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Spuig, Pascal; Le-luyer, Alain; Malard, Philippe; Cantone, Bruno; Pastor, Patrick; Saint-Laurent, François [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Vayakis, George; Delhom, Dominique [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Arshad, Shakeib [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Lister, Jonathan; Toussaint, Matthieu; Marmillod, Philippe; Testa, Duccio; Schlatter, Christian [Ecole polytechnique fédérale de Lausanne, Centre de Recherches en Physique des Plasmas, 1015 Lausanne (Switzerland); Peruzzo, Simone [Consorzio RFX, C.so Stati Uniti 4, 35127 Padova (Italy)

2015-10-15

Highlights: • ITER Continuous External Rogowskis are designed for plasma current measurement. • CER are located in the casing of Toroidal Field Coils and will operate at 4.5 K. • The design of the sensors has been completed and validated through prototypes. • Detailed assembly procedure inside the toroidal field coil casing has been defined. • The CER has passed all the ITER and F4E design review procedures. - Abstract: In ITER, an accurate measurement of plasma current, with high reliability, is mandatory as this parameter is used to demonstrate licensing compliance with regulatory limits. For that purpose, several independent measurements based on magnetic diagnostics have been proposed. Rogowski coils are standard inductive sensors for current measurement in many applications. In ITER, three continuous external Rogowski coils are to be installed in the casing of the toroidal field coils. These sensors are remarkable from several points of view: overall length is about 40 m, high sensitivity needed, located in the toroidal field coil casing at 4.5 K and complex 3D routing with tight bending radius of 50 mm. Since 2005 an extensive work has been carried out to develop and analyze several design options complying with ITER specifications. Prototypes of a selected continuous external Rogowski design were built and tested successfully in terms of electrical, thermal, mechanical and vacuum characteristics. Finally a detailed assembly procedure inside the toroidal field coil casing has been defined according to the coil manufacturing and assembly constraints.

ITER CTA newsletter. No. 13, October 2002

International Nuclear Information System (INIS)

2002-11-01

This ITER CTA newsletter issue comprises concise information about an ITER related meeting concerning the joint implementation of ITER - the fifth ITER Negotiations Meeting - which was held in Toronto, Canada, 19-20 September, 2002, and information about assessment of the possible ITER site in Clarington, Ontario, Canada, which was the subject of the first official stage of the Joint Assessment of Specific Sites (JASS) for the ITER Project. This assessment was completed just before the Fifth ITER Negotiations Meeting

The ITER remote maintenance system

International Nuclear Information System (INIS)

Tesini, A.; Palmer, J.

2008-01-01

The aim of this paper is to summarize the ITER approach to machine components maintenance. A major objective of the ITER project is to demonstrate that a future power producing fusion device can be maintained effectively and offer practical levels of plant availability. During its operational lifetime, many systems of the ITER machine will require maintenance and modification; this can be achieved using remote handling methods. The need for timely, safe and effective remote operations on a machine as complex as ITER and within one of the world's most hostile remote handling environments represents a major challenge at every level of the ITER Project organization, engineering and technology. The basic principles of fusion reactor maintenance are presented. An updated description of the ITER remote maintenance system is provided. This includes the maintenance equipment used inside the vacuum vessel, inside the hot cell and the hot cell itself. The correlation between the functions of the remote handling equipment, of the hot cell and of the radwaste processing system is also described. The paper concludes that ITER has equipped itself with a good platform to tackle the challenges presented by its own maintenance and upgrade needs

Design earthquakes for ITER in Europe at Cadarache

International Nuclear Information System (INIS)

Girard, Jean-Philippe; Gruenthal, Gottfried; Nicolas, Marc

2005-01-01

The European site proposed for ITER is situated in the south of France, 40 km north-east of Aix-en-Provence, in a low to moderate seismic area according to the Global Seismic Hazard Map (GSHAP Group 1999). The tokamak building would be implemented on good bedrock made of limestone with a shear wave velocity of over 1300 m/s. Input requirements and assumptions for ITER consider that an infrequent, severe earthquake (called SL-2), although unlikely to occur during the lifetime of the facility, is assessed to demonstrate adequate protection of the public. This earthquake is assumed to have a return period of 10,000 years. An investment protection level or inspection level (where all structures, systems and components are safe) with a peak ground acceleration (pga) at 0.5 m/s 2 is also considered. As a basis, orders of magnitude of consequences, if no countermeasures were taken, are given. Four aspects are discussed: regulation, implementation of this regulation for the proposed site (site geology, tectonic and seismotectonic), a probabilistic seismic hazard assessment of the site and finally, the fulfilment of the requirements and assumptions, according to IAEA guides. As a conclusion of the studies, the main characteristics of the Cadarache European site are discussed. Preliminary studies have shown that the European site proposal will ensure a low level of project risk with respect to the seismic hazard

The development and preliminary testing of a multimedia patient–provider survivorship communication module for breast cancer survivors

Science.gov (United States)

Wen, Kuang-Yi; Miller, Suzanne M.; Stanton, Annette L.; Fleisher, Linda; Morra, Marion E.; Jorge, Alexandra; Diefenbach, Michael A.; Ropka, Mary E.; Marcus, Alfred C.

2012-01-01

Objective This paper describes the development of a theory-guided and evidence-based multimedia training module to facilitate breast cancer survivors’ preparedness for effective communication with their health care providers after active treatment. Methods The iterative developmental process used included: (1) theory and evidence-based content development and vetting; (2) user testing; (3) usability testing; and (4) participant module utilization. Results Formative evaluation of the training module prototype occurred through user testing (n = 12), resulting in modification of the content and layout. Usability testing (n = 10) was employed to improve module functionality. Preliminary web usage data (n = 256, mean age = 53, 94.5% White, 75% college graduate and above) showed that 59% of the participants accessed the communication module, for an average of 7 min per login. Conclusion The iterative developmental process was informative in enhancing the relevance of the communication module. Preliminary web usage results demonstrate the potential feasibility of such a program. Practice implications Our study demonstrates survivors’ openness to the use of a web-based communication skills training module and outlines a systematic iterative user and interface program development and testing process, which can serve as a prototype for others considering such an approach. PMID:22770812

The development and preliminary testing of a multimedia patient-provider survivorship communication module for breast cancer survivors.

Science.gov (United States)

Wen, Kuang-Yi; Miller, Suzanne M; Stanton, Annette L; Fleisher, Linda; Morra, Marion E; Jorge, Alexandra; Diefenbach, Michael A; Ropka, Mary E; Marcus, Alfred C

2012-08-01

This paper describes the development of a theory-guided and evidence-based multimedia training module to facilitate breast cancer survivors' preparedness for effective communication with their health care providers after active treatment. The iterative developmental process used included: (1) theory and evidence-based content development and vetting; (2) user testing; (3) usability testing; and (4) participant module utilization. Formative evaluation of the training module prototype occurred through user testing (n = 12), resulting in modification of the content and layout. Usability testing (n = 10) was employed to improve module functionality. Preliminary web usage data (n = 256, mean age = 53, 94.5% White, 75% college graduate and above) showed that 59% of the participants accessed the communication module, for an average of 7 min per login. The iterative developmental process was informative in enhancing the relevance of the communication module. Preliminary web usage results demonstrate the potential feasibility of such a program. Our study demonstrates survivors' openness to the use of a web-based communication skills training module and outlines a systematic iterative user and interface program development and testing process, which can serve as a prototype for others considering such an approach. Copyright © 2012. Published by Elsevier Ireland Ltd.

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