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Sample records for iter vacuum systems

  1. ITER diagnostic system: Vacuum interface

    Energy Technology Data Exchange (ETDEWEB)

    Patel, K.M., E-mail: Kaushal.Patel@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Drevon, J.M. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Encheva, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Kashchuk, Y. [Institution “PROJECT CENTER ITER”, 1, Akademika Kurchatova pl., Moscow (Russian Federation); Maquet, Ph. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France)

    2013-10-15

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10{sup −7} Pa, irrespective of plasma operation, and a leak rate of less than 10{sup −10} Pa m{sup 3} s{sup −1}. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions

  2. ITER diagnostic system: Vacuum interface

    International Nuclear Information System (INIS)

    Patel, K.M.; Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L.; Drevon, J.M.; Encheva, A.; Kashchuk, Y.; Maquet, Ph.; Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J.

    2013-01-01

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10 −7 Pa, irrespective of plasma operation, and a leak rate of less than 10 −10 Pa m 3 s −1 . In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions will be described

  3. 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.

  4. Microwave response of ITER vacuum windows

    NARCIS (Netherlands)

    Oosterbeek, J.W.; Maquet, P.; Sirinelli, A.; Udintsev, V.S.; Vayakis, G.; Walsh, M.J.

    2017-01-01

    Diagnostic systems are essential for the development of ITER discharges and to reach the ITER goals. Many of these diagnostics require a line of sight to relay signals from the plasma to the diagnostic, typically located outside the torus hall. Such diagnostics then require vacuum windows that

  5. Validated design of the ITER main vacuum pumping systems

    International Nuclear Information System (INIS)

    Day, Chr.; Antipenkov, A.; Dremel, M.; Haas, H.; Hauer, V.; Mack, A.; Boissin, J.-C.; Class, G.; Murdoch, D.K.; Wykes, M.

    2005-01-01

    Forschungszentrum Karlsruhe is developing the ITER high vacuum cryogenic pumping systems (torus, cryostat, NBI) as well as the corresponding mechanical roughing pump trains. All force-cooled big cryopumps incorporate similar design of charcoal coated cryopanels cooled to 5 K with supercritical helium. A model of the torus exhaust cryopump was comprehensively characterised in the TIMO testbed at Forschungszentrum. This paper discusses the vacuum performance results of the model pump and outlines how these data were incorporated in a sound design of the whole ITER torus exhaust pumping system. To do this, the dedicated software package ITERVAC was developed which is able to describe gas flow in viscous, transitional and molecular flow regimes as needed for the gas coming through the divertor slots and along the pump ducts into the cryopumps. The entrance section between the divertor cassettes and each pumping duct was identified to be the bottleneck of the gas flow. The interrelation of achievable throughputs as a function of the divertor pressure and the cryopump pumping speed is discussed. The system design is completed by assessment of the NBI cryopump system and integrating performance curves for the roughing pump trains needed during the regeneration phases of the cryopumps. (author)

  6. Vacuum Bellows, Vacuum Piping, Cryogenic Break, and Copper Joint Failure Rate Estimates for ITER Design Use

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader

    2010-06-01

    The ITER international project design teams are working to produce an engineering design in preparation for construction of the International Thermonuclear Experimental Reactor (ITER) tokamak. During the course of this work, questions have arisen in regard to safety barriers and equipment reliability as important facets of system design. The vacuum system designers have asked several questions about the reliability of vacuum bellows and vacuum piping. The vessel design team has asked about the reliability of electrical breaks and copper-copper joints used in cryogenic piping. Research into operating experiences of similar equipment has been performed to determine representative failure rates for these components. The following chapters give the research results and the findings for vacuum system bellows, power plant stainless steel piping (amended to represent vacuum system piping), cryogenic system electrical insulating breaks, and copper joints.

  7. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

    Stringer, J.

    1992-11-01

    This report describes further design issues concerning remote maintenance of torus vacuum pumping systems options for ITER. The key issues under investigation in this report are flask support systems for valve seal exchange operations for the compound cryopump scheme and remote maintenance of a proposed multiple turbomolecular pump (TMP) system, an alternative ITER torus exhaust pumping option. Previous studies have shown that the overhead support methods for seal exchange flask equipment could malfunction due to valve/flask misalignment. A floor-mounted support system is described in this report. This scheme provides a more rigid support system for seal exchange operations. An alternative torus pumping system, based on the use of multiple TMPs, is studied from a remote maintenance standpoint. In this concept, centre distance spacing for pump/valve assemblies is too restrictive for remote maintenance. Recommendations are made for adequate spacing of these assemblies based on commercially-available 0.8 m and 1.0 m diameter valves. Fewer pumps will fit in this arrangement, which implies a need for larger TMPs. Pumps of this size are not commercially available. Other concerns regarding the servicing and storage of remote handling equipment in cells are also identified. (9 figs.)

  8. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    International Nuclear Information System (INIS)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10 -4 Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  9. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  10. Modeling of complex gas distribution systems operating under any vacuum conditions: Simulations of the ITER divertor pumping system

    International Nuclear Information System (INIS)

    Vasileiadis, N.; Tatsios, G.; Misdanitis, S.; Valougeorgis, D.

    2016-01-01

    Highlights: • An integrated s/w for modeling complex rarefied gas distribution systems is presented. • Analysis is based on kinetic theory of gases. • Code effectiveness is demonstrated by simulating the ITER divertor pumping system. • The present s/w has the potential to support design work in large vacuum systems. - Abstract: An integrated software tool for modeling and simulation of complex gas distribution systems operating under any vacuum conditions is presented and validated. The algorithm structure includes (a) the input geometrical and operational data of the network, (b) the definition of the fundamental set of network loops and pseudoloops, (c) the formulation and solution of the mass and energy conservation equations, (d) the kinetic data base of the flow rates for channels of any length in the whole range of the Knudsen number, supporting, in an explicit manner, the solution of the conservation equations and (e) the network output data (mainly node pressures and channel flow rates/conductance). The code validity is benchmarked under rough vacuum conditions by comparison with hydrodynamic solutions in the slip regime. Then, its feasibility, effectiveness and potential are demonstrated by simulating the ITER torus vacuum system with the six direct pumps based on the 2012 design of the ITER divertor. Detailed results of the flow patterns and paths in the cassettes, in the gaps between the cassettes and along the divertor ring, as well as of the total throughput for various pumping scenarios and dome pressures are provided. A comparison with previous results available in the literature is included.

  11. Progress and Achievements on the R&D Activities for ITER Vacuum Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, M. [Japan Atomic Energy Research Institute (JAERI); Koizumi, K. [Japan Atomic Energy Research Institute (JAERI); Takahashi, H. [Japan Atomic Energy Research Institute (JAERI); Onozuka, M. [ITER Joint Central Team, Garching, Germany; Ioki, K. [ITER Joint Central Team, Garching, Germany; Kuzumin, E. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Krylov, V. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Maslakowski, J. [Oak Ridge National Laboratory (ORNL); Nelson, Brad E [ORNL; Jones, L. [Max-Planck Institute, Garching, Germany; Danner, W. [Max-Planck Institute, Garching, Germany; Maisonnier, D. [Max-Planck Institute, Garching, Germany

    2001-01-01

    The ITER vacuum vessel (VV) is designed to be large double-walled structure with a D-shaped crosssection. The achievable fabrication tolerance of this structure was unknown due to the size and complexity of shape. The Full-scale Sector Model of ITER Vacuum Vessel, which was 15m in height, was fabricated and tested to obtain the fabrication and assembly tolerances. The model was fabricated within the target tolerance of 5mm and welding deformation during assembly operation was obtained. The port structure was also connected using remotized welding tools to demonstrate the basic maintenance activity. In parallel, the tests of advanced welding, cutting and inspection system were performed to improve the efficiency of fabrication and maintenance of the Vacuum Vessel. These activities show the feasibility of ITER Vacuum Vessel as feasible in a realistic way. This paper describes the major progress, achievement and latest status of the R&D activities on the ITER vacuum vessel.

  12. Study of ITER equatorial port plug handling system and vacuum sealing interface

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Jean-Pierre [Association Euratom CEA, CEA/DSM/IRFM, Cadarache, F-13108 Saint-Paul-lez-Durance (France)], E-mail: jean-pierre.martins@cea.fr; Doceul, Louis; Marol, Sebastien; Delchie, Elise [Association Euratom CEA, CEA/DSM/IRFM, Cadarache, F-13108 Saint-Paul-lez-Durance (France); Cordier, Jean-Jacques; Levesy, Bruno; Tesini, Alessandro [ITER International Organization, F-13108 Saint-Paul-lez-Durance cedex (France); Ciattaglia, Emanuela [EFDA CSU - Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Tivey, Richard [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Gillier, Rene; Abbes, Christophe [GARLOCK - Sealing Technologies - 90, rue de la roche du Geai, F-42029 St-Etienne cedex 1 (France)

    2009-06-15

    In the field of the ITER port plug engineering and integration task, CEA has contributed to define proposals concerning the port plugs vacuum sealing interface with the vessel flange and the equatorial plug handling. The 2001 baseline vacuum flange sealing consisted of TIG welding of a 316L strip plate on to U shapes. This arrangement presented some issues like welding access, implementation of tools, lip consumption, complex local leak test, continuous leak checking. Therefore, an alternate sealing solution based on the use of metallic gaskets is proposed. The different technical aspects are discussed to explain how this design can simplify the maintenance and deal with safety and vacuum requirements. The design of the mechanical attachment and vacuum sealing of the plug has constantly evolved, but the associated remote handling equipment was not systematically reviewed. An update of the cask and maintenance procedure was studied in order to design it in accordance with the last generic plug flange design. This includes a concept of a gripping system that uses the plug flange bolting area and, to help the remote handling process, a cantilever assisting system is suggested to increase the reliability of the transfer operation between vacuum vessel and cask.

  13. 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

  14. 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

  15. Lubricant coating of dowel for the ITER vacuum vessel gravity support

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.Y. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Ahn, H.J., E-mail: hjahn@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bak, J.S. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Choi, C.H.; Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Zauner, C. [KRP-Mechatec Engineering GbR, 85748 Garching b, Muenchen (Germany)

    2012-08-15

    The ITER vacuum vessel gravity supports located in the lower level shall sustain loads in radial, toroidal and vertical directions. The hinge type VVGS consists of two hinges, upper and lower blocks and dowels. In order to develop the design concept and verify the structural integrity of the hinge system, the design analysis has been performed in detail. Inclination of 15 Degree-Sign for the hinge based supporting system was introduced to provide centering force to make stable equilibrium state of the vacuum vessel. Due to this inclination the hinges are rotated by the radial expansion of the VV during operation and baking, respectively. If a dowel is seized in the hinge, the supporting system can be highly stressed due to the restrained displacement in the seized dowel. Therefore, solid lubricant coatings were suggested on dowels in order to avoid seizing in the sliding area. In this work, several sets of coupons were made with different coating materials to investigate the effect according to the selection of coating material. Also, a test facility was designed to cover the ITER relevant loading and boundary conditions, e.g. vacuum condition, temperature, contact pressure, cycles, etc. From those test results, the optimized coating method was found to avoid seizure of dowel in the ITER VVGS.

  16. Fabrication progress of the ITER vacuum vessel sector in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.C., E-mail: bckim@nfri.re.kr [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Lee, Y.J.; Hong, K.H.; Sa, J.W.; Kim, H.S.; Park, C.K.; Ahn, H.J.; Bak, J.S.; Jung, K.J. [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Park, K.H.; Roh, B.R.; Kim, T.S.; Lee, J.S.; Jung, Y.H.; Sung, H.J.; Choi, S.Y.; Kim, H.G.; Kwon, I.K.; Kwon, T.H. [Hyundai Heavy Industries Co. Ltd., Dong-gu, Ulsan (Korea, Republic of)

    2013-10-15

    Highlights: ► Fabrication of ITER vacuum vessel sector full scale mock-up to develop fabrication procedures. ► The welding and nondestructive examination techniques conform to RCC-MR. ► The preparation of real manufacturing of ITER vacuum vessel sector. -- Abstract: As a participant of ITER project, ITER Korea has to supply two ITER vacuum vessel sectors (Sector no. 6, no. 1) of total nine ITER VV sectors. After the procurement arrangement with ITER Organization, ITER Korea made the contract with Hyundai Heavy Industries (HHI) for fabrication of two sectors. Then the start of the manufacturing design was initiated from January 2010. HHI made three real scale R and D mock-ups to verify the critical fabrication feasibility issues on electron beam welding, 3D forming, welding distortion and achievable tolerances. The documentation according to IO and the French nuclear safety regulation requirement, the qualification of welding and nondestructive examination procedures conform to RCC-MR 2007 were proceed in parallel. The mass production of raw material was done after receiving ANB (agreed notified body) verification of product/parts and shop qualification. The manufacturing drawing, manufacturing and inspection plan of VV sector with supporting fabrication procedures are also verified by ANB, accordingly the first cutting and forming of plates for VV sector fabrication started from February 2012. This paper reports the latest fabrication progress of ITER vacuum vessel Sector no. 6 that will be assembled as the first sector in the ITER pit. The overall fabrication route, R and D mock-up fabrication results with forming and welding distortion analysis, qualification status of welding and nondestructive examination (NDE) are also presented.

  17. Status of the ITER vacuum vessel construction

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C.H.; Sborchia, C.; Ioki, K.; Giraud, B.; Utin, Yu.; Sa, J.W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Wang, X., E-mail: xiaoyuwww@gmail.com [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Teissier, P.; Martinez, J.M.; Le Barbier, R.; Jun, C.; Dani, S.; Barabash, V.; Vertongen, P.; Alekseev, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Jucker, P.; Bayon, A. [F4E, c/ Josep Pla, n. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Pathak, H.; Raval, J. [ITER-India, IPR, A-29, Electronics Estate, GIDC, Sector-25, Gandhinagar 382025 (India); Ahn, H.J. [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); and others

    2014-10-15

    Highlights: • Final design of the ITER vacuum vessel (VV). • Procurement of the ITER VV. • Manufacturing results of real scale mock-ups. • Manufacturing status of the VV in domestic agencies. - Abstract: The ITER vacuum vessel (VV) is under manufacturing by four domestic agencies after completion of engineering designs that have been approved by the Agreed Notified Body (ANB). Manufacturing designs of the VV have been being completed, component by component, by accommodating requirements of the RCC-MR 2007 edition. Manufacturing of the VV first sector has been started in February 2012 in Korea and in-wall shielding in May 2013 in India. EU will start manufacturing of its first sector from September 2013 and Russia the upper port by the end of 2013. All DAs have manufactured several mock-ups including real-size ones to justify/qualify and establish manufacturing techniques and procedures.

  18. Vacuum vessel port structures for ITER-FEAT

    International Nuclear Information System (INIS)

    Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th.

    2001-01-01

    The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed

  19. Vacuum vessel port structures for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th

    2001-11-01

    The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed.

  20. Integration of ITER in-vessel diagnostic components in the vacuum vessel

    International Nuclear Information System (INIS)

    Encheva, A.; Bertalot, L.; Macklin, B.; Vayakis, G.; Walker, C.

    2009-01-01

    The integration of ITER in-vessel diagnostic components is an important engineering activity. The positioning of the diagnostic components must correlate not only with their functional specifications but also with the design of the major parts of ITER torus, in particular the vacuum vessel, blanket modules, blanket manifolds, divertor, and port plugs, some of which are not yet finally designed. Moreover, the recently introduced Edge Localised Mode (ELM)/Vertical Stability (VS) coils mounted on the vacuum vessel inner wall call for not only more than a simple review of the engineering design settled down for several years now, but also for a change in the in-vessel distribution of the diagnostic components and their full impact has yet to be determined. Meanwhile, the procurement arrangement (a document defining roles and responsibilities of ITER Organization and Domestic Agency(s) (DAs) for each in-kind procurement including technical scope of work, quality assurance requirements, schedule, administrative matters) for the vacuum vessel must be finalized. These make the interface process even more challenging in terms of meeting the vacuum vessel (VV) procurement arrangement's deadline. The process of planning the installation of all the ITER diagnostics and integrating their installation into the ITER Integrated Project Schedule (IPS) is now underway. This paper covers the progress made recently on updating and issuing the interfaces of the in-vessel diagnostic components with the vacuum vessel, outlines the requirements for their attachment and summarises the installation sequence.

  1. ITER vacuum vessel structural analysis completion during manufacturing phase

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F. [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Caixas, J.; Fernandez, E.; Zarzalejos, J.M. [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Kim, H.-S.; Kim, Y.G. [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Privalova, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); and others

    2016-11-01

    Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

  2. ITER vacuum vessel structural analysis completion during manufacturing phase

    International Nuclear Information System (INIS)

    Martinez, J.-M.; Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F.; Caixas, J.; Fernandez, E.; Zarzalejos, J.M.; Kim, H.-S.; Kim, Y.G.; Privalova, E.

    2016-01-01

    Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

  3. Clearance potential of ITER vacuum vessel activated materials

    International Nuclear Information System (INIS)

    Cepraga, D.G.; Cambi, G.; Frisoni, M.

    2002-01-01

    To demonstrate fusion's environmental attractiveness over the entire life cycle, a waste analysis is mandatory. The clearance is recommended by IAEA for releasing activated solid materials from regulatory control and for waste management policy. The paper focuses on the approach used to support waste analyses for ITER Generic Site Safety Report. The Material Unconditional Clearance Index of all the materials/zones on the equatorial mid-plane of ITER machine have been evaluated, based on IAEA-TECDOC-855. The Bonami-Nitawl-XSDNRPM sequence of the Scale-4.4a code system (using Vitenea-J library) has been firstly used for radiation transport analyses. Then the Anita-2000 code package is used for the activation calculation. The paper presents also, as an example, an application of the clearance indexes estimation for the ITER vacuum vessel materials. The results of the Anita-2000 have been compared with those obtained using the Fispact-99 activation code. (author)

  4. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

    Stringer, J.

    1992-11-01

    This report describes design issues concerning remote maintenance of the ITER torus vacuum pumping system. Key issues under investigation in this report are bearings for inert gas operation, transporter integration options, cryopump access, gate valve maintenance frequency, tritium effects on materials, turbomolecular pump design, and remote maintenance. Alternative bearing materials are explored for inert gas operation. Encapsulated motors and rotary feedthroughs offer an alternative option where space requirements are restrictive. A number of transporter options are studied. The preferred scheme depends on the shielded reconfigured ducts to prevent streaming and activation of RH (remote handling) equipment. A radiation mapping of the cell is required to evaluate this concept. Valve seal and bellow life are critical issues and need to be evaluated, as they have a direct bearing on the provision of adequate RH equipment to meet scheduled and unscheduled maintenance outages. The limited space on the inboard side of the cryopumps for RH equipment access requires a reconfigured duct and manifold. A modified shielded duct arrangement is proposed, which would provide more access space, reduced activation of components, and the potential for improved valve seal life. Work at Mound Laboratories has shown the adverse effects of tritium on some bearing lubricants. Silicone-based lubricants should be avoided. (11 refs., 2 tabs., 31 figs.)

  5. ITER articulated inspection arm (AIA): R and d progress on vacuum and temperature technology for remote handling

    International Nuclear Information System (INIS)

    Perrot, Y.; Cordier, J.J.; Friconneau, J.P.; Gargiulo, L.; Martin, E.; Palmer, J.D.; Tesini, A.

    2005-01-01

    This paper is part of the remote handling (RH) activities for the future fusion reactor ITER. The aim of the R and D program performed under the European Fusion Development Agreement (EFDA) work program is to demonstrate the feasibility of close inspection tasks such as viewing or leak testing of the Divertor cassettes and the Vacuum Vessel (VV) first wall of ITER. It is assumed that a long reach, limited payload carrier penetrates the ITER chamber through the openings evenly distributed around the machine such as In-Vessel Viewing System (IVVS) access or through upper port plugs. To perform an intervention a short time after plasma shut down, the operation of the robot should be realised under ITER conditioning i.e. under high vacuum and temperature conditions (120 o C). The feasibility analysis drove the design of the so-called articulated inspection arm (AIA) which is a 8.2 m long robot made of five modules with a 11 actuated joints kinematics. A single module prototype was designed in detail and manufactured to be tested under ITER realistic conditions at CEA-Cadarache test facility. As well as demonstrating the potential for the application of an AIA type device in ITER, this program is also dedicated to explore the necessary robotic technologies required to ITER's IVVS deployment system. This paper presents the whole AIA robot concept, the first results of the test campaign on the prototype vacuum and temperature demonstrator module

  6. ITER articulated inspection arm (AIA): R and d progress on vacuum and temperature technology for remote handling

    Energy Technology Data Exchange (ETDEWEB)

    Perrot, Y. [Robotics and Interactive Systems Unit-CEA/LIST, BP6 F-92265 Fontenay aux Roses Cedex (France)]. E-mail: yann.perrot@cea.fr; Cordier, J.J. [DRFC-CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Friconneau, J.P. [Robotics and Interactive Systems Unit-CEA/LIST, BP6 F-92265 Fontenay aux Roses Cedex (France); Gargiulo, L. [DRFC-CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Martin, E. [ITER International Team, Boltzmannstrasse 2, 85748 Garching (Germany); Palmer, J.D. [EFDA CSU Garching, Boltzmannstrasse 2, 85748 Garching (Germany); Tesini, A. [ITER International Team, ITER Naka Joint Work Site, 801-1, Muouyama, Naka-machi, Naka-gun, Iberaki-ken 311-0193 (Japan)

    2005-11-15

    This paper is part of the remote handling (RH) activities for the future fusion reactor ITER. The aim of the R and D program performed under the European Fusion Development Agreement (EFDA) work program is to demonstrate the feasibility of close inspection tasks such as viewing or leak testing of the Divertor cassettes and the Vacuum Vessel (VV) first wall of ITER. It is assumed that a long reach, limited payload carrier penetrates the ITER chamber through the openings evenly distributed around the machine such as In-Vessel Viewing System (IVVS) access or through upper port plugs. To perform an intervention a short time after plasma shut down, the operation of the robot should be realised under ITER conditioning i.e. under high vacuum and temperature conditions (120 {sup o}C). The feasibility analysis drove the design of the so-called articulated inspection arm (AIA) which is a 8.2 m long robot made of five modules with a 11 actuated joints kinematics. A single module prototype was designed in detail and manufactured to be tested under ITER realistic conditions at CEA-Cadarache test facility. As well as demonstrating the potential for the application of an AIA type device in ITER, this program is also dedicated to explore the necessary robotic technologies required to ITER's IVVS deployment system. This paper presents the whole AIA robot concept, the first results of the test campaign on the prototype vacuum and temperature demonstrator module.

  7. Remote maintenance of a combined regeneration-isolation valve for the ITER Torus vacuum pumping system

    International Nuclear Information System (INIS)

    Stringer, J.; Blevins, J.

    1992-01-01

    A large diameter valve suitable for high vacuum operation is under study for ITER Torus evacuation. The valves must comply with specifications for leak-tightness, radiation resistance, dust tolerance, overpressure, and thermal gradients. Remote maintenance of the seal and valve moving parts without disturbance to the rest of the valve system is a requirement. This paper describes tow methods of seal exchange by remote means. In the first method, a flask is proposed for the valve moving parts exchange in inert gas, when the machine is shut down. In the second method a novel concept is described for seal exchange while under vacuum, without having to bring the machine up to atmosphere. The advantages of this method are that scheduled remote handling (RH) operations and outages for seal replacement are not required. Also, the need for a flask is avoided

  8. Assembly of the sectors and ports of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Corino, S.; Moreno, R.

    2014-01-01

    The International Thermonuclear Experimental Reactor, ITER is a very complex Project that aims to prove the technical reliability of nuclear fusion. ITER has been Ensa's commitment to the future to strengthen as one of the main manufacturers of big equipment and services internationally in the nuclear field. Ensa started working on the qualification process to be able to bid for the 'Assembly of the ITER vacuum vessel' in June 2010, after two and a half years of pre-qualification, offers, clarifications and long technical meetings, that were followed by commercial meetings Ensa achieved its goal. The 30 of November, Ensa signed what at that time was the biggest of the supplies signed by IO (ITER Organization). A lot of efforts and hard work had been done in order to achieve this goal, but the hardest of all was yet to come, after the signature of the contract, Ensa has 7 years ahead to achieve the final goal, the assembly and welding of the 9 sectors that put together the ITER vacuum vessel and the 54 ports that will allow the assembly of the different auxiliary systems. The scope of the works to be performed can generally be divided into the following areas: - Welding of the sectors and ports; - Non-destructive tests; - Machining; - Dimensional Controls. In order to achieve this goal, the project has been divided into 3 different phases. - Development phase: January 2013 - July 2015; - Pre-production phase: July 2015 - February 2016; - Production phase: February 2016 - February 2020

  9. ITER vacuum vessel, in vessel components and plasma facing materials

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Enoeda, M.; Federici, G.

    2007-01-01

    in a protected position flush with the FW. There are no sliding supports inside the vacuum, to keep the reliability of the system. Driving mechanisms are located outside the vacuum boundary. The divertor activities have progressed with the aim of launching the procurement according to the ITER project schedule. They include: (a) the consolidation of the design and manufacturing technologies for the plasma facing components (PFCs); (b) the prequalification programme by the parties prior to entering into the procurement phase, (c) the diagnostics integration into the divertor design, (d) the development of suitable acceptance criteria for the divertor PFCs including the required fabrication control methods; (e) the development of remote handling procedures for the first installation and for the following replacements of the divertor cassettes. (orig.)

  10. Progress and achievements of R&D activities for the ITER vacuum vessel

    Science.gov (United States)

    Nakahira, M.; Takahashi, H.; Koizumi, K.; Onozuka, M.; Ioki, K.

    2001-04-01

    The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R&D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18° toroidal sector, which is composed of two 9° sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of +/-3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R&D activities for the ITER vacuum vessel.

  11. Development of an inspection robot under iter relevant vacuum and temperature conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hatchressian, J-C; Bruno, V; Gargiulo, L; Bayetti, P; Cordier, J-J; Samaille, F [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA Cadarache, F-13108 Saint Paul-Lez-Durance Cedex (France); Keller, D; Perrot, Y; Friconneau, J-P [CEA, LIST, Service de Robotique Interactive, 18 route du Panorama, BP6, Fontenay aux Roses F-92265 France (France); Palmer, J D [EFDA-CSU Max-Planck-Institut fuer Plasma Physik Boltzmannstr.2, D-85748 Garching Germany (Germany)

    2008-03-15

    Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In vessel inspection operations without loss of conditioning could be very mandatory. Within this framework, the aim of the Articulated Inspection Arm (AIA) project is to demonstrate the feasibility of a multi-purpose in-vessel Remote Handling inspection system. It is a long reach, composed of 5 segments with in all 8 degrees of freedom, limited payload carrier (up to 10kg) and a total range of 8m. The project is currently developed by the CEA within the European work program. Some tests will validate chosen concepts for operations under ITER relevant vacuum and temperature conditions. The presence of magnetic fields, radiation and neutron beams will not be considered. This paper deals with the choices of the materials to minimize the out-gassing under vacuum and high temperature during conditioning, the implantation of the electronics which are enclosed in boxes with special gaskets, the design of the first embedded process which is a viewing system.

  12. The ITER remote maintenance system

    International Nuclear Information System (INIS)

    Tesini, A.; Palmer, J.

    2007-01-01

    ITER is a joint international research and development project that aims to demonstrate the scientific and technological feasibility of fusion power. As soon as the plasma operation begins using tritium, the replacement of the vacuum vessel internal components will need to be done with remote handling techniques. To accomplish these operations ITER has equipped itself with a Remote Maintenance System; this includes the Remote Handling equipment set and the Hot Cell facility. Both need to work in a cooperative way, with the aim of minimizing the machine shutdown periods and to maximize the machine availability. The ITER Remote Handling equipment set is required to be available, robust, reliable and retrievable. The machine components, to be remotely handle-able, are required to be designed simply so as to ease their maintenance. The baseline ITER Remote Handling equipment is described. The ITER Hot Cell Facility is required to provide a controlled and shielded area for the execution of repair operations (carried out using dedicated remote handling equipment) on those activated components which need to be returned to service, inside the vacuum vessel. The Hot Cell provides also the equipment and space for the processing and temporary storage of the operational and decommissioning radwaste. A conceptual ITER Hot Cell Facility is described. (orig.)

  13. IWR-solution for the ITER vacuum vessel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Wu, H., E-mail: huapeng@lut.fi [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Handroos, H. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Pela, P. [Tekes (Finland); Wang, Y. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland)

    2011-10-15

    The assembly of ITER vacuum vessel (VV) is still a very big challenge as the process can only be done from inside the VV. The welding of the VV assembly is carried out using the dedicated robotic systems. The main functions of the robots are: (i) measuring the actual space between every two sectors, (ii) positioning of the 150 kg splice plates between the sector shells, (iii) welding the splice plates to the sector shells, (iv) NDT of the welds, (v) repairing, including machining of the welds, (vi) He-leak tests of the welds, and (vii) the non-planned functions that may turn out. This paper presents a reasonable method to assemble the ITER VV. In this article, one parallel mobile robot, running on the track rail fixed on the wall inside the VV, is designed and tested. The assembling process, carried out by the mobile robot together with the welding robot, is presented.

  14. Design of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

    1995-01-01

    The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement.The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2-2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4-0.7 m.The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200-300 MN due to VDE and to allow for differential thermal expansion.The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology. (orig.)

  15. Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

    Science.gov (United States)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R.

    1998-10-01

    Design and R&D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R&D results. The resulting design changes are discussed for each system.

  16. Design and development of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Koizumi, K.; Nakahira, M.; Itou, Y.; Tada, E. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan); Johnson, G.; Ioki, K.; Elio, F.; Iizuka, T.; Sannazzaro, G.; Takahashi, K.; Utin, Y.; Onozuka, M. [ITER Joint Central Team (JCT), Garching (Germany); Nelson, B. [US Home Team, Oak Ridge National Laboratory (United States); Vallone, C. [EU Home Team, NET Team, Garching (Germany); Kuzmin, E. [RF Home Team, Efremov Institute, City (Russian Federation)

    1998-09-01

    In ITER, the vacuum vessel (VV) is designed to be a water cooled, double-walled toroidal structure made of 316LN stainless steel with a D-shaped cross section approximately 9 m wide and 15 m high. The design work which began at the beginning of the ITER-EDA is nearing completion by resolving the technical issues. In parallel with the design activities, the R and D program, full-scale VV sector model project, was initiated in 1995 to resolve the design and fabrication issues. The full-scale sector model corresponds to an 18 sector (9 sub-sector x 2) and is being fabricated on schedule. To date, 60% of the fabrication had been completed. The fabrication of full-scale model including sector-to-sector connection will be completed by the end of 1997 and performance tests are scheduled until the end of ITER-EDA. This paper describes the latest status of the ITER VV design and the full-scale sector model project. (orig.) 3 refs.

  17. 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

  18. Progress and achievements of R and D activities for the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Nakahira, M.; Takahashi, H.; Koizumi, K.; Onozuka, M.; Ioki, K.

    2001-01-01

    The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R and D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18 deg. toroidal sector, which is composed of two 9 deg. sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of ±3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R and D activities for the ITER vacuum vessel. (author)

  19. Integration test of ITER full-scale vacuum vessel sector

    International Nuclear Information System (INIS)

    Nakahira, M.; Koizumi, K.; Oka, K.

    1999-01-01

    The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven ITER R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of - 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (author)

  20. Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R. [ITER JCT, Garching (Germany)

    1998-10-01

    Design and R and D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R and D results. The resulting design changes are discussed for each system. (orig.) 11 refs.

  1. A mobile robot with parallel kinematics to meet the requirements for assembling and machining the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Pessi, Pekka [Lappeenranta University of Technology, Lappeenranta (Finland)], E-mail: pessi@lut.fi; Wu, Huapeng; Handroos, Heikki [Lappeenranta University of Technology, Lappeenranta (Finland); Jones, Lawrence [EFDA Close Support Unit, Boltzmannstrasse 2, Garching D-85748 (Germany)

    2007-10-15

    The present paper introduces a mobile parallel robot developed for International Thermonuclear Experimental Reactor (ITER). The task of the robot is to carry out welding and machining processes inside the ITER vacuum vessel. The kinematic design of the robot has been optimized for the ITER access. The kinematic analysis is given in the paper. A virtual prototype of the parallel robot is built. A dynamic behavior of the whole robot is studied by the multi-body system simulation (MBS)

  2. A mobile robot with parallel kinematics to meet the requirements for assembling and machining the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Pessi, Pekka; Wu, Huapeng; Handroos, Heikki; Jones, Lawrence

    2007-01-01

    The present paper introduces a mobile parallel robot developed for International Thermonuclear Experimental Reactor (ITER). The task of the robot is to carry out welding and machining processes inside the ITER vacuum vessel. The kinematic design of the robot has been optimized for the ITER access. The kinematic analysis is given in the paper. A virtual prototype of the parallel robot is built. A dynamic behavior of the whole robot is studied by the multi-body system simulation (MBS)

  3. Integration test of ITER full-scale vacuum vessel sector

    International Nuclear Information System (INIS)

    Nakahira, M.; Koizumi, K.; Oka, K.

    2001-01-01

    The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of ± 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (author)

  4. 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)

  5. 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 systemvacuum 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

  6. Modeling and analysis of alternative concept of ITER vacuum vessel primary heat transfer system

    International Nuclear Information System (INIS)

    Carbajo, Juan; Yoder, Graydon; Dell'Orco, G.; Curd, Warren; Kim, Seokho

    2010-01-01

    A RELAP5-3D model of the ITER (Latin for 'the way') vacuum vessel (VV) primary heat transfer system has been developed to evaluate a proposed design change that relocates the heat exchangers (HXs) from the exterior of the tokamak building to the interior. This alternative design protects the HXs from external hazards such as wind, tornado, and aircraft crash. The proposed design integrates the VV HXs into a VV pressure suppression system (VVPSS) tank that contains water to condense vapour in case of a leak into the plasma chamber. The proposal is to also use this water as the ultimate sink when removing decay heat from the VV system. The RELAP5-3D model has been run under normal operating and abnormal (decay heat) conditions. Results indicate that this alternative design is feasible, with no effects on the VVPSS tank under normal operation and with tank temperature and pressure increasing under decay heat conditions resulting in a requirement to remove steam generated if the VVPSS tank low pressure must be maintained.

  7. ITER vacuum vessel fabrication plan and cost study (D 68) for the international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    1995-01-01

    ITER Task No. 8, Vacuum Vessel Fabrication Plan and Cost Study (D68), was initiated to assess ITER vacuum vessel fabrication, assembly, and cost. The industrial team of Raytheon Engineers ampersand Constructors and Chicago Bridge ampersand Iron (Raytheon/CB ampersand I) reviewed the current vessel basis and prepared a manufacturing plan, assembly plan, and cost estimate commensurate with the present design. The guidance for the Raytheon/CB ampersand I assessment activities was prepared by the ITER Garching Work Site. This guidance provided in the form of work descriptions, sketches, drawings, and costing guidelines for each of the presently identified vacuum vessel Work Breakdown Structure (WBS) elements was compiled in ITER Garching Joint Work Site Memo (Draft No. 9 - G 15 MD 01 94-17-05 W 1). A copy of this document is provided as Appendix 1 to this report. Additional information and clarifications required for the Raytheon/CB ampersand I assessments were coordinated through the US Home Team (USHT) and its technical representative. Design details considered essential to the Task 8 assessments but not available from the ITER Joint Central Team (JCT) were generated by Raytheon/CB ampersand I and documented accordingly

  8. FW/Blanket and vacuum vessel for RTO/RC ITER

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M.

    2000-01-01

    The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste

  9. FW/Blanket and vacuum vessel for RTO/RC ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M

    2000-11-01

    The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, {approx}50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste.

  10. Alternatives of ITER vacuum vessel support system

    International Nuclear Information System (INIS)

    Ohmori, Junji; Kitamura, Kazunori; Araki, Masanori; Ohno, Isamu; Shoji, Teruaki

    2002-07-01

    Optional designs of vacuum vessel (VV) support have been performed for the International Thermonuclear Experimental Reactor (ITER) to reduce stresses and buckling concern of the flexible plate structure in ITER-FDR. One of the optional designs is hanging type VV support concept that consists of top hanging supports at the top of VV and middle radial stoppers in the middle of outboard VV. The hanging supports are located at the toroidal field (TF) coil inboard top region (R∼5400 mm) using the narrow window space surrounded by a poloidal field coil (PF1) and TF coil. The radial stoppers are located inside TF coil cases in the TF coil outboard middle region (R∼9300 mm). The upper flange connection of the radial stoppers should slide in vertical direction to eliminate thermal stress produced by relative thermal displacement between VV wall and TF coil case. Both supports consist of flexible plates and are mounted on 18 locations in toroidal direction. The radial and toroidal reaction forces are shared with the hanging supports and the radial stoppers. However, the vertical force is sustained by only the hanging supports. The others are compressive type support concept that consists of nine VV supports located in alternate divertor port regions in toroidal direction. Two designs have been performed for the VV support concept. One is mounted on TF inter-coil structures (OIS) and the other is on cryostat ring. The compressive support on TF coil OIS is dependent on TF coil movement but that on cryostat is independent. In the optional designs, the bending stress due to the relative thermal displacement between TF coil and VV is classified to primary stress according to ASME Sec. III NF. The stress due to TF coil displacement is also considered as primary stress. The stress due to non-uniform temperature distribution of the flexible plate is classified to secondary stress. The preliminary structural assessments for the optional designs have been performed for all load

  11. Progress of ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K., E-mail: Kimihiro.Ioki@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bayon, A. [F4E, c/ Josep Pla, No. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Kim, B.C. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Kuzmin, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Le Barbier, R.; Martinez, J.-M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Pathak, H. [ITER-India, A-29, GIDC Electronic Estate, Sector 25, Gandhinagar 382025 (India); Preble, J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sa, J.W. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Terasawa, A.; Utin, Yu. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); and others

    2013-10-15

    Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure.

  12. Progress of ITER vacuum vessel

    International Nuclear Information System (INIS)

    Ioki, K.; Bayon, A.; Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B.; Kim, B.C.; Kuzmin, E.; Le Barbier, R.; Martinez, J.-M.; Pathak, H.; Preble, J.; Sa, J.W.; Terasawa, A.; Utin, Yu.

    2013-01-01

    Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure

  13. 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.)

  14. Development of tritium fuel processing system using electrolytic reactor for ITER

    International Nuclear Information System (INIS)

    Yamanishi, T.; Kawamura, Y.; Iwai, Y.

    2001-01-01

    The system composed of a palladium diffuser and an electrolytic reactor was proposed, and was developed for a Fuel Cleanup system of ITER. The performance of the system was studied in a stand-alone test in detail. A fuel simulation loop of ITER was constructed by connecting the developed Fuel Cleanup and Hydrogen Isotope Separation systems; and the function of each system in the loop was demonstrated. For the tritium recovery from the exhaust gas at He glow discharge cleaning of vacuum chamber of ITER, a cryogenic molecular sieve bed system was proposed and demonstrated. (author)

  15. Development of tritium fuel processing system using electrolytic reactor for ITER

    International Nuclear Information System (INIS)

    Yamanishi, Toshihiko; Kawamura, Y.; Iwai, Y.

    1999-01-01

    The system composed of a palladium diffuser and an electrolytic reactor was proposed, and was developed for a Fuel Cleanup system of ITER. The performance of the system was studied in a stand-alone test in detail. A fuel simulation loop of ITER was constructed by connecting the developed Fuel Cleanup and Hydrogen Isotope Separation systems; and the function of each system in the loop was demonstrated. For the tritium recovery from the exhaust gas at He glow discharge cleaning of vacuum chamber of ITER, a cryogenic molecular sieve bed system was proposed and demonstrated. (author)

  16. Automation of Aditya vacuum control system based on CODAC Core System

    Energy Technology Data Exchange (ETDEWEB)

    Raulji, Vismaysinh D., E-mail: vismay@ipr.res.in; Pujara, Harshad; Arambhadiya, Bharat; Jadeja, Kumarpalsinh; Bhatt, Shailesh; Rajpal, Rachana

    2016-11-15

    Highlights: • Monitor and control of vacuum control system based on CODAC Core System. • Communication between SIEMENS PLC and open source software EPICS. • With CODAC Core easy to configure and programming of slow controller. - Abstract: The main objective of vacuum control system is to provide ultrahigh vacuum for Aditya Tokamak operations. Aditya Vacuum vessel is having four vacuum pumping lines. To demonstrate implementation of automation; a study case is under taken by automating single Pumping Line of the Aditya vacuum system using CODAC Core System (CCS). Currently, vacuum system is operated manually. The CCS based control system allows remote control, monitoring, alarm handling of vacuum parameters. The CODAC Core System is the Linux based software package that is distributed by ITER Organization for the development of Plant System I&C software. CODAC Core System includes EPICS, CSS (Control System Studio) etc. CSS is used for HMI (Human Machine Interface), alarms and archives. SDD (Self Description Data) tool is used to configure plant system I&C. SDD Editor is an Eclipse based application to define the plant system, interface, I&C component, interfaced signals, configure variable. SCADA (Supervisory Control and Data Acquisition) system is developed in CSS. Data is transferred between PLC and CSS through EPICS. The complete system is tested with Aditya Vacuum Control System with process interlocks. Operator interface is also developed using Lab VIEW as a choice of the user. This paper will describe the salient features of the developed control system in detail.

  17. A mobile robot with parallel kinematics constructed under requirements for assembling and machining of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Pessi, P.; Huapeng Wu; Handroos, H.; Jones, L.

    2006-01-01

    ITER sectors require more stringent tolerances ± 5 mm than normally expected for the size of structure involved. The walls of ITER sectors are made of 60 mm thick stainless steel and are joined together by high efficiency structural and leak tight welds. In addition to the initial vacuum vessel assembly, sectors may have to be replaced for repair. Since commercially available machines are too heavy for the required machining operations and the lifting of a possible e-beam gun column system, and conventional robots lack the stiffness and accuracy in such machining condition, a new flexible, lightweight and mobile robotic machine is being considered. For the assembly of the ITER vacuum vessel sector, precise positioning of welding end-effectors, at some distance in a confined space from the available supports, will be required, which is not possible using conventional machines or robots. This paper presents a special robot, able to carry out welding and machining processes from inside the ITER vacuum vessel, consisting of a ten-degree-of-freedom parallel robot mounted on a carriage driven by electric motor/gearbox on a track. The robot consists of a Stewart platform based parallel mechanism. Water hydraulic cylinders are used as actuators to reach six degrees of freedom for parallel construction. Two linear and two rotational motions are used for enlargement the workspace of the manipulator. The robot carries both welding gun such as a TIG, hybrid laser or e-beam welding gun to weld the inner and outer walls of the ITER vacuum vessel sectors and machining tools to cut and milling the walls with necessary accuracy, it can also carry other tools and material to a required position inside the vacuum vessel . For assembling an on line six degrees of freedom seam finding algorithm has been developed, which enables the robot to find welding seam automatically in a very complex environment. In the machining multi flexible machining processes carried out automatically by

  18. Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Shibui, Masanao; Koizumi, Koichi; Kanamori, Naokazu; Nishio, Satoshi; Sasaki, Takashi; Tada, Eisuke

    1992-09-01

    Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

  19. The ITER tritium systems

    International Nuclear Information System (INIS)

    Glugla, M.; Antipenkov, A.; Beloglazov, S.; Caldwell-Nichols, C.; Cristescu, I.R.; Cristescu, I.; Day, C.; Doerr, L.; Girard, J.-P.; Tada, E.

    2007-01-01

    ITER is the first fusion machine fully designed for operation with equimolar deuterium-tritium mixtures. The tokamak vessel will be fuelled through gas puffing and pellet injection, and the Neutral Beam heating system will introduce deuterium into the machine. Employing deuterium and tritium as fusion fuel will cause alpha heating of the plasma and will eventually provide energy. Due to the small burn-up fraction in the vacuum vessel a closed deuterium-tritium loop is required, along with all the auxiliary systems necessary for the safe handling of tritium. The ITER inner fuel cycle systems are designed to process considerable and unprecedented deuterium-tritium flow rates with high flexibility and reliability. High decontamination factors for effluent and release streams and low tritium inventories in all systems are needed to minimize chronic and accidental emissions. A multiple barrier concept assures the confinement of tritium within its respective processing components; atmosphere and vent detritiation systems are essential elements in this concept. Not only the interfaces between the primary fuel cycle systems - being procured through different Participant Teams - but also those to confinement systems such as Atmosphere Detritiation or those to fuelling and pumping - again procured through different Participant Teams - and interfaces to buildings are calling for definition and for detailed analysis to assure proper design integration. Considering the complexity of the ITER Tritium Plant configuration management and interface control will be a challenging task

  20. Design of parallel intersector weld/cut robot for machining processes in ITER vacuum vessel

    International Nuclear Information System (INIS)

    Wu Huapeng; Handroos, Heikki; Kovanen, Janne; Rouvinen, Asko; Hannukainen, Petri; Saira, Tanja; Jones, Lawrence

    2003-01-01

    This paper presents a new parallel robot Penta-WH, which has five degrees of freedom driven by hydraulic cylinders. The manipulator has a large, singularity-free workspace and high stiffness and it acts as a transport device for welding, machining and inspection end-effectors inside the ITER vacuum vessel. The presented kinematic structure of a parallel robot is particularly suitable for the ITER environment. Analysis of the machining process for ITER, such as the machining methods and forces are given, and the kinematic analyses, such as workspace and force capacity are discussed

  1. 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)

  2. Manufacturing and maintenance technologies developed for a thick-wall structure of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Onozuka, M.; Alfile, J.P.; Aubert, Ph.; Dagenais, J.-F.; Grebennikov, D.; Ioki, K.; Jones, L.; Koizumi, K.; Krylov, V.; Maslakowski, J.; Nakahira, M.; Nelson, B.; Punshon, C.; Roy, O.; Schreck, G.

    2001-01-01

    Development of welding, cutting and non-destructive testing (NDT) techniques, and development of remotized systems have been carried out for on-site manufacturing and maintenance of the thick-wall structure of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV). Conventional techniques, including tungsten inert gas welding, plasma cutting, and ultrasonic inspection, have been improved and optimized for the application to thick austenitic stainless steel plates. In addition, advanced methods have been investigated, including reduced-pressure electron-beam and multi-pass neodymium-doped yttrium aluminum garnet (NdYAG) laser welding, NdYAG laser cutting, and electro-magnetic acoustic transducer inspection, to improve cost and technical performance. Two types of remotized systems with different payloads have been investigated and one of them has been fabricated and demonstrated in field joint welding, cutting, and NDT tests on test mockups and full-scale ITER VV sector models. The progress and results of this development to date provide a high level of confidence that the manufacturing and maintenance of the ITER VV is feasible

  3. Manufacturing and maintenance technologies developed for a thick-wall structure of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. E-mail: onozukm@itereu.de; Alfile, J.P.; Aubert, Ph.; Dagenais, J.-F.; Grebennikov, D.; Ioki, K.; Jones, L.; Koizumi, K.; Krylov, V.; Maslakowski, J.; Nakahira, M.; Nelson, B.; Punshon, C.; Roy, O.; Schreck, G

    2001-09-01

    Development of welding, cutting and non-destructive testing (NDT) techniques, and development of remotized systems have been carried out for on-site manufacturing and maintenance of the thick-wall structure of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV). Conventional techniques, including tungsten inert gas welding, plasma cutting, and ultrasonic inspection, have been improved and optimized for the application to thick austenitic stainless steel plates. In addition, advanced methods have been investigated, including reduced-pressure electron-beam and multi-pass neodymium-doped yttrium aluminum garnet (NdYAG) laser welding, NdYAG laser cutting, and electro-magnetic acoustic transducer inspection, to improve cost and technical performance. Two types of remotized systems with different payloads have been investigated and one of them has been fabricated and demonstrated in field joint welding, cutting, and NDT tests on test mockups and full-scale ITER VV sector models. The progress and results of this development to date provide a high level of confidence that the manufacturing and maintenance of the ITER VV is feasible.

  4. 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

  5. Simulation of LLCB TBM in-vessel first wall coolant break into ITER vacuum vessel by using RELAP/MOD3.4

    International Nuclear Information System (INIS)

    Tony Sandeep, K.; Chaudhari, Vilas; Rajendra Kumar, E.; Dutta, Anu; Singh, R.K.

    2013-06-01

    To prove Test Blanket Module (TBM) safety in International Thermonuclear Experimental Reactor (ITER), various accident scenarios are postulated . One of the postulated initiating events to be analysed is TBM First wall (FW) coolant leak in ITER Vacuum vessel (VV). This accident has been classified as a reference event for the TBM (probability of occurrence >1 E -06 /a). The postulated accident occurs as a result of small leak of TBM FW helium into ITER vacuum vessel (VV), caused by the TBM weld failure. The ingress of this TBM FW helium into ITER plasma induces intense plasma disruption that deposits 1.8 MJ/m 2 of plasma stored thermal energy onto the TBM FW over a period of 1 sec in duration (assumption). Runaway electrons in this process are lost from plasma current channel and cause multiple TBM and ITER FW cooling tube failures within 10 cm torriodal strip. The size of the break is identified as double ended rupture of all coolant channels within this strip around the reactor. For LLCB TBM this represents failure of 4 FW channels. The size of ITER FW break is 0.02 m 2 . Consequently, a simultaneous blow down of TBM FW helium and ITER FW water occurs, injecting helium and water into VV. This pressurisation causes the activation of VV pressure suppressions system and ingress of water into VV. This pressurisation causes the VV pressure suppressions system (VVPSS) to open in an attempt to contain the pressure below the safety limit of 0.2 MPa. This report is intended to do the above accident analysis and assessment of active components of TBM using RELAP code and hence prove its safety in ITER environment. (author)

  6. ITER fuel cycle

    International Nuclear Information System (INIS)

    Leger, D.; Dinner, P.; Yoshida, H.

    1991-01-01

    Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

  7. Demonstration tests for manufacturing the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Onozuka, Masanori; Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro; Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi

    2007-01-01

    Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV

  8. Demonstration tests for manufacturing the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Katsusuke [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan)], E-mail: katsusuke_shimizu@mhi.co.jp; Onozuka, Masanori [Mitsubishi Heavy Industries, Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan); Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)

    2007-10-15

    Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV.

  9. Parameter study on Japanese proposal of ITER hydrogen isotope separation system

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Enoeda, Mikio; Tanaka, Shigeru; Ohokawa, Yoshinao; Ohara, Atsushi; Nagakura, Masaaki; Naito, Taisei; Nagashima, Kazuhiro.

    1991-01-01

    As part of Japanese design contribution in the ITER activity, conceptual design of an entire ITER tritium system and their safety analysis have been carried out through the three-year period since 1988. The tritium system includes the following subsystems; - Fuelling (gas puffing and pellet injection) subsystem, - Torus vacuum pumping subsystem, - Plasma exhaust gas purification subsystem, - Hydrogen isotope separation subsystem, - NBI gas processing subsystem, - Blanket tritium recovery subsystem, - Tritiated water processing subsystem, - Tritium safety subsystem. Hydrogen isotope separation system is a key subsystem in the ITER tritium system because it is connected to all above subsystems. This report describes an analytical study on the Japanese concept of hydrogen isotope separation system. (author)

  10. Structural analysis of the ITER vacuum vessel from disruption loading with halo asymmetry

    International Nuclear Information System (INIS)

    Riemer, B.W.; Sayer, R.O.

    1996-01-01

    Static structural analyses of the ITER vacuum vessel were performed with toroidally asymmetric disruption loads. Asymmetric halo current conditions were assumed to modify symmetric disruption loads which resulted in net lateral loading on the vacuum vessel torus. Structural analyses with the asymmetric loading indicated significantly higher vessel stress and blanket support forces than with symmetric disruption loads. A recent change in the vessel support design which provided toroidal constraints at each mid port was found to be effective in reducing torus lateral movement and vessel stress

  11. Robot vision system R and D for ITER blanket remote-handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Tesini, Alessandro

    2014-01-01

    For regular maintenance of the International Thermonuclear Experimental Reactor (ITER), a system called the ITER blanket remote-handling system is necessary to remotely handle the blanket modules because of the high levels of gamma radiation. Modules will be handled by robotic power manipulators and they must have a non-contact-sensing system for installing and grasping to avoid contact with other modules. A robot vision system that uses cameras was adopted for this non-contact-sensing system. Experiments for grasping modules were carried out in a dark room to simulate the environment inside the vacuum vessel and the robot vision system's measurement errors were studied. As a result, the accuracy of the manipulator's movements was within 2.01 mm and 0.31°, which satisfies the system requirements. Therefore, it was concluded that this robot vision system is suitable for the non-contact-sensing system of the ITER blanket remote-handling system

  12. Robot vision system R and D for ITER blanket remote-handling system

    Energy Technology Data Exchange (ETDEWEB)

    Maruyama, Takahito, E-mail: maruyama.takahito@jaea.go.jp [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Tesini, Alessandro [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France)

    2014-10-15

    For regular maintenance of the International Thermonuclear Experimental Reactor (ITER), a system called the ITER blanket remote-handling system is necessary to remotely handle the blanket modules because of the high levels of gamma radiation. Modules will be handled by robotic power manipulators and they must have a non-contact-sensing system for installing and grasping to avoid contact with other modules. A robot vision system that uses cameras was adopted for this non-contact-sensing system. Experiments for grasping modules were carried out in a dark room to simulate the environment inside the vacuum vessel and the robot vision system's measurement errors were studied. As a result, the accuracy of the manipulator's movements was within 2.01 mm and 0.31°, which satisfies the system requirements. Therefore, it was concluded that this robot vision system is suitable for the non-contact-sensing system of the ITER blanket remote-handling system.

  13. Manufacturing preparations for the European Vacuum Vessel Sector for ITER

    International Nuclear Information System (INIS)

    Jones, Lawrence; Arbogast, Jean François; Bayon, Angel; Bianchi, Aldo; Caixas, Joan; Facca, Aldo; Fachin, Gianbattista; Fernández, José; Giraud, Benoit; Losasso, Marcello; Löwer, Thorsten; Micó, Gonzalo; Pacheco, Jose Miguel; Paoletti, Roberto; Sanguinetti, Gian Paolo; Stamos, Vassilis; Tacconelli, Massimiliano; Trentea, Alexandru; Utin, Yuri

    2012-01-01

    The contract for the seven European Sectors of the ITER Vacuum Vessel, which has very tight tolerances and high density of welding, was placed at the end of 2010 with AMW, a consortium of three companies. The start-up of the engineering, including R and D, design and analysis activities of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. The statutory and regulatory requirements of ITER Organization and the French Nuclear Safety regulations have made the design development subject to rigorous controls. AMW was able to make use of the previous extensive R and D and prototype work carried out during the past 9 years, especially in relation to advanced welding and inspection techniques. The paper describes the manufacturing methodology with the focus on controlling distortion with predictions by analysis, avoiding use of welded-on jigs, and making use of low heat input narrow-gap welding with electron beam welding as far as possible and narrow-gap TIG when not. Further R and D and more than ten significant mock-ups are described. All these preparations will help to assure the successful manufacture of this critical path item of ITER.

  14. 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.

  15. Structural analysis of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Sannazzaro, G.; Ioki, K.; Johnson, G.; Onozuka, M.; Utin, Y. [ITER Joint Work Site, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States); Swanson, J. [USHT, Raytheon, Princeton (United States)

    1998-07-01

    The ITER Vacuum Vessel (VV) must withstand a large number of loading conditions including electromagnetic, seismic, operational and upset pressure, thermal and test loads. All of the loading conditions and load combinations have been categorized and classified to permit the allowable stress to be defined in accordance with the recommendations of the ASME code. The most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and the load combination of seismic and electromagnetic loads due to a plasma vertical instability. The areas of high stress are the regions around the VV and the blanket supports, and the attachment of the ports to the main shell. In all of the loading conditions and load combinations the calculated stresses are below the allowable values. (authors)

  16. Design and integration of lower ports for ITER diagnostic systems

    Energy Technology Data Exchange (ETDEWEB)

    Casal, Natalia, E-mail: Natalia.casal@iter.org [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); Bertalot, Luciano; Cheng, Hao; Drevon, Jean Marc; Duckworth, Philip; Giacomin, Thibaud; Guirao, Julio; Iglesias, Silvia [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); Kochergin, Mikhail [IOFFE Institute, Saint Petersburg (Russian Federation); Martin, Alex [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); McCarron, Eddie [Oxford Technologies Ltd., Abingdon (United Kingdom); Mokeev, Alexander [Russian Federation Domestic Agency, Moscow (Russian Federation); Mota, Fernando [CIEMAT, Madrid (Spain); Penot, Christophe; Portales, Mickael [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); Kitazawa, Sin-iti [Japanese Domestic Agency, Naka (Japan); Sky, Jack [Oxford Technologies Ltd., Abingdon (United Kingdom); Suarez, Alejandro; Udintsev, Victor; Utin, Yuri [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); and others

    2015-10-15

    Highlights: • Lower port structures are in its conceptual design phase. • Electromagnetic and seismic loads, will dominate all other mechanical loads. • Design allows diagnostics support, neutron shielding while and signals transmission. • Installation and maintenance operations are fully remote handling compatible. - Abstract: All around the ITER vacuum vessel, forty-four ports will provide access to the vacuum vessel for remote handling operations, diagnostic systems, heating, and vacuum systems: 18 upper ports, 17 equatorial ports, and 9 lower ports. Among the lower ports, three of them will be used for the remote handling installation of the ITER divertor. Once the divertor is in place, these ports will host various diagnostic systems mounted in the so-called diagnostic racks. The diagnostic racks must allow the support and cooling of the diagnostics, extraction of the required diagnostic signals, and providing access and maintainability while minimizing the leakage of radiation toward the back of the port where the humans are allowed to enter. A fully integrated inner rack, carrying the near plasma diagnostic components, will be an stainless steel structure, 4.2 m long, with a maximum weight of 10 t. This structure brings water for cooling and baking at maximum temperature of 240 °C and provides connection with gas, vacuum and electric services. Additional racks (placed away from plasma and not requiring cooling) may be required for the support of some particular diagnostic components. The diagnostics racks and its associated ex vessel structures, which are in its conceptual design phase, are being designed to survive the lifetime of ITER of 20 years. This paper presents the current state of development including interfaces, diagnostic integration, operation and maintenance, shielding requirements, remote handling, loads cases and discussion of the main challenges coming from the severe environment and engineering requirements.

  17. 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

  18. 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.

  19. Hydrogen/hydrocarbon explosions in the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Goranson, P.L.

    1992-01-01

    The consequences of H 2 /hydrocarbon detonations in the vacuum vessel (torus) of the International Thermonuclear Experimental Reactor (ITER) have been studied. The most likely scenario for such a detonation involves a water leak into the torus and a vent of the torus to atmosphere, permitting the formation of an explosive fuel-air mixture. The generation of fuel gases and possible sources of air or oxygen are reviewed, and the severity and effects of specific fuel-air mixture explosions are evaluated. Detonation or deflagration of an explosive mixture could result in pressures exceeding the maximum allowable torus pressure. Further studies to examine the design details and develop an event-tree study of events following a gas detonation are recommended

  20. 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

  1. Dielectric window development for the ITER ICRF vacuum transmission line

    International Nuclear Information System (INIS)

    Heikinheimo, L.; Heikkinen, J.; Hytoenen, Y.

    1998-08-01

    A vacuum window block design is presented for the ITER (International Thermonuclear Experimental Reactor) ion cyclotron radio frequency heating vacuum transmission line. The vacuum windows in various auxiliaries of the present fusion facilities and in future fusion reactors are essential and most vulnerable components, as they provide ultimate vacuum and tritium containment. Various existing windows, e.g. those at the ASDEX Upgrade tokamak (IPP, Garching), Tore Supra (CEA, Cadarache), and at JET Tokamak (Euratom, Culham), have provided a starting point for the present work, but the large size, remote handling and cooling requirements, as well as the strong neutron radiation in the fusion reactors have called a new design, where new material combinations for the dielectric and conductor have deemed to be necessary. Conventional and well tested design solutions have been chosen wherever possible. A preprototype construction has been launched to test the design principles, in particular the joining of the ceramics to a titanium conductor by brazing which has not been possible so far in this size and geometry. First results of the preprototype tests are reported here. The present final report is a combination of the results achieved with the support of two different projects; an industrial project funded by EFET and a NET Research and Design project funded by the Euratom/TEKES within the Euratom fusion programme. (orig.)

  2. Materials requirements for the ITER vacuum vessel and in-vessel components - approaching the construction phase

    International Nuclear Information System (INIS)

    Barabash, V.; Ioki, K.; Pick, M.; Girard, J.P.; Merola, M.

    2007-01-01

    Full text of publication follows: The ITER activities are fully devoted toward its construction. In accordance with the ITER integrated project schedule, the procurement specifications for the manufacturing of the Vacuum Vessel should be prepared by March 2008 and the procurement specifications for the in-vessel components (first wall/blanket, divertor) by 2009. To update the design, considering design and technology evolution, the ITER Design Review has been launched. Among the various topics being discussed are the important issues related to selection of materials, material procurement, and assessment of performance during operation. The main requirements related to materials for the vacuum vessel and the in-vessel components are summarized in the paper. The specific licensing requirements are to be followed for structural materials of pressure and nuclear pressure equipment components for construction of ITER. In addition, the procurements in ITER will be done mostly 'in-kind' and it is assumed that materials for these components will be produced by different Parties. However, in accordance with the regulatory requirements and quality requirements for operation, common specifications and the general rules to fulfill these requirements are to be adopted. For some ITER components (e.g. first wall, divertor high heat flux components), the ultimate qualification of the joining technologies (Be/Cu, SS/Cu, CFC/Cu, W/Cu) is under final evaluation. Successful accomplishment of the qualification program will allow to proceed with procurements of the components for ITER. The criteria for acceptance of these components and materials after manufacturing are described and the main results will be reported. Additional materials issues, which come from the on-going manufacturing R and D program, will be also described. Finally, further materials activity during the construction phase, needs for final qualification and acceptance of materials are discussed. (authors)

  3. Systematic vacuum study of the ITER model cryopump by test particle Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Xueli; Haas, Horst; Day, Christian [Institute for Technical Physics, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2011-07-01

    The primary pumping systems on the ITER torus are based on eight tailor-made cryogenic pumps because not any standard commercial vacuum pump can meet the ITER working criteria. This kind of cryopump can provide high pumping speed, especially for light gases, by the cryosorption on activated charcoal at 4.5 K. In this paper we will present the systematic Monte Carlo simulation results of the model pump in a reduced scale by ProVac3D, a new Test Particle Monte Carlo simulation program developed by KIT. The simulation model has included the most important mechanical structures such as sixteen cryogenic panels working at 4.5 K, the 80 K radiation shield envelope with baffles, the pump housing, inlet valve and the TIMO (Test facility for the ITER Model Pump) test facility. Three typical gas species, i.e., deuterium, protium and helium are simulated. The pumping characteristics have been obtained. The result is in good agreement with the experiment data up to the gas throughput of 1000 sccm, which marks the limit for free molecular flow. This means that ProVac3D is a useful tool in the design of the prototype cryopump of ITER. Meanwhile, the capture factors at different critical positions are calculated. They can be used as the important input parameters for a follow-up Direct Simulation Monte Carlo (DSMC) simulation for higher gas throughput.

  4. Dynamic simulation of relief line during loss of insulation vacuum of the ITER cryoline

    Science.gov (United States)

    Badgujar, S.; Kosek, J.; Grillot, D.; Forgeas, A.; Sarkar, B.; Shah, N.; Choukekar, K.; Chang, H.-S.

    2017-12-01

    The ITER cryoline (CL) system consists of 37 types of vacuum jacketed transfer lines which forms a complex structured network with a total length of about 5 km, spread inside the Tokamak building, on a dedicated plant bridge and in the Cryoplant building/area. One of them, the low pressure relief line (RL) recovers helium discharged from process safety relief valves of the different cryogenic users and is sent it back to the Cryoplant via heater and recovery system. The process pipe diameters of the RL vary from DN 50 to DN 200 and the length is more than 1500 m. Loss of insulation vacuum (LIV) of a CL is one of the worst scenarios apart from LIV in Auxiliary Cold Boxes (ACBs). The Torus and Cryostat CL is chosen to simulate the virtual LIV and to study the anticipated behavior of the RL. Both helium LIV (LIV due to leak in helium pipe) and air LIV (LIV due to air ingress in outer vacuum jacket of the cryoline) with and without fire) have been simulated during this study. After the brief description of the CL system, the paper will describe the EcosimPro® model prepared for the dynamic study. The paper will also describe the results like minimum temperature of RL, mass flow and maximum pressure in the RL which are essentially used to choose the type and location of safety relief devices to protect the CL process pipes.

  5. Advanced Examination Techniques Applied to the Assessment of Vacuum Pressure Impregnation (VPI) of ITER Correction Coils

    CERN Document Server

    Sgobba, Stefano; Samain, Valerie; Libeyre, Paul; Cecillon, Alexandre; Dawid, J

    2014-01-01

    The ITER Magnet System includes a set of 18 superconducting correction coils (CC) which are used to compensate the error field modes arising from geometrical deviations caused by manufacturing and assembly tolerances. The turn and ground insulation are electrically insulated with a multi-layer fiberglass polyimide interleaved composite, impregnated with epoxy resin using vacuum pressure impregnation (VPI). Adequate high voltage insulation (5 kV), mechanical strength and rigidity of the winding pack should be achieved after impregnation and curing of the insulation system. VPI is an effective process to avoid defects such dry spots and incomplete wet out. This insulation technology has also been developed since several years for application to large superconducting coils and more recently to ITER CC. It allows the coils to be impregnated without impacting on their functional characteristics. One of the critical challenges associated with the construction of the CC is the qualification of the VPI insulation. Se...

  6. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    Science.gov (United States)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

    2000-12-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ˜50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R&D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R&D is being performed.

  7. General and crevice corrosion study of the in-wall shielding materials for ITER vacuum vessel

    Science.gov (United States)

    Joshi, K. S.; Pathak, H. A.; Dayal, R. K.; Bafna, V. K.; Kimihiro, Ioki; Barabash, V.

    2012-11-01

    Vacuum vessel In-Wall Shield (IWS) will be inserted between the inner and outer shells of the ITER vacuum vessel. The behaviour of IWS in the vacuum vessel especially concerning the susceptibility to crevice of shielding block assemblies could cause rapid and extensive corrosion attacks. Even galvanic corrosion may be due to different metals in same electrolyte. IWS blocks are not accessible until life of the machine after closing of vacuum vessel. Hence, it is necessary to study the susceptibility of IWS materials to general corrosion and crevice corrosion under operations of ITER vacuum vessel. Corrosion properties of IWS materials were studied by using (i) Immersion technique and (ii) Electro-chemical Polarization techniques. All the sample materials were subjected to a series of examinations before and after immersion test, like Loss/Gain weight measurement, SEM analysis, and Optical stereo microscopy, measurement of surface profile and hardness of materials. After immersion test, SS 304B4 and SS 304B7 showed slight weight gain which indicate oxide layer formation on the surface of coupons. The SS 430 material showed negligible weight loss which indicates mild general corrosion effect. On visual observation with SEM and Metallography, all material showed pitting corrosion attack. All sample materials were subjected to series of measurements like Open Circuit potential, Cyclic polarization, Pitting potential, protection potential, Critical anodic current and SEM examination. All materials show pitting loop in OC2 operating condition. However, its absence in OC1 operating condition clearly indicates the activity of chloride ion to penetrate oxide layer on the sample surface, at higher temperature. The critical pitting temperature of all samples remains between 100° and 200°C.

  8. Development of ITER in-vessel viewing and metrology systems

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation ({approx}30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  9. Development of ITER in-vessel viewing and metrology systems

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira

    1998-01-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation (∼30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  10. Structural analysis of support structure for ITER vacuum vessel

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Ohmori, Junji; Nakahira, Masataka

    2004-12-01

    ITER vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed. This independent concept has two advantages: (1) thermal load due to the temperature deference between VV and the lower temperature components such as TF coil becomes lower and (2) the other components such as TF coil is categorized as a non-safety component because of its independence from VV. Stress analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support. (author)

  11. Design progress of the ITER vacuum vessel sectors and port structures

    International Nuclear Information System (INIS)

    Utin, Yu.; Ioki, K.; Alekseev, A.; Bachmann, Ch.; Cho, S.; Chuyanov, V.; Jones, L.; Kuzmin, E.; Morimoto, M.; Nakahira, M.; Sannazzaro, G.

    2007-01-01

    Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on achievements of manufacturing studies, design improvement of the typical VV Sector (no. 1) has been nearly finalized. Design improvement of other sectors is in progress-in particular, of the VV Sectors no. 2 and no. 3 which interface with tangential ports for the neutral beam (NB) injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. In particular, design of the NB ports was advanced with the focus on the beam-facing components to handle the heat input of the neutral beams. Structural analyses have been performed to validate all design improvements

  12. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

    2000-01-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ∼50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed

  13. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.deiokik@ipp.mpg.de; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M

    2000-12-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve {approx}50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed.

  14. MEA vacuum system

    International Nuclear Information System (INIS)

    Stroo, R.; Schwebke, H.; Heine, E.

    1984-01-01

    This report describes construction and operation of the MEA vacuum system of NIKHEF (Netherlands). First, the klystron vacuum system, beam transport system, diode pump and a triode pump are described. Next, the isolation valve and the fast valves of the vacuum system are considered. Measuring instruments, vacuum system commands and messages of failures are treated in the last chapter. (G.J.P.)

  15. 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.

  16. Simulation of VDE under intervention of vertical stability control and vertical electromagnetic force on the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Miyamoto, S.; Sugihara, M.; Shinya, K.; Nakamura, Y.; Toshimitsu, S.; Lukash, V.E.; Khayrutdinov, R.R.; Sugie, T.; Kusama, Y.; Yoshino, R.

    2012-01-01

    Highlights: ► Taking account of intervention of VS control, VDE simulations were carried out. ► Malfunctioning of VS circuit (positive feedback) enhances the vertical force. ► The worst case was explored for vertical force on the ITER vacuum vessel. ► We confirmed the force is still within the design margin even if the worst case. - Abstract: Vertical displacement events (VDEs) and disruptions usually take place under intervention of vertical stability (VS) control and the vertical electromagnetic force induced on vacuum vessels is potentially influenced. This paper presents assessment of the force that arises from the VS control in ITER VDEs using a numerical simulation code DINA. The focus is on a possible malfunctioning of the ex-vessel VS control circuit: radial magnetic field is unintentionally applied to the direction of enhancing the vertical displacement further. Since this type of failure usually causes the largest forces (or halo currents) observed in the present experiments, this situation must be properly accommodated in the design of the ITER vacuum vessel. DINA analysis shows that although the ex-vessel VS control modifies radial field, it does not affect plasma motion and current quench behavior including halo current generation because the vacuum vessel shields the field created by the ex-vessel coils. Nevertheless, the VS control modifies the force on the vessel by directly acting on the eddy current carried by the conducting structures of the vessel. Although the worst case was explored in a range of plasma inductance and pattern of VS control in combination with the in-vessel VS control circuit, the result confirmed that the force is still within the design margin.

  17. Design and analysis of the vacuum vessel for RTO/RC-ITER

    International Nuclear Information System (INIS)

    Onozuka, M.; Ioki, K.; Johnson, G.; Kodama, T.; Sannazzaro, G.; Utin, Y.

    2000-01-01

    Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a 'tight fitting' configuration is required to correctly position the modules' first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible

  18. Design and analysis of the vacuum vessel for RTO/RC-ITER

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Johnson, G.; Kodama, T.; Sannazzaro, G.; Utin, Y

    2000-11-01

    Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a 'tight fitting' configuration is required to correctly position the modules' first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible.

  19. Tests of dry mechanical forepumps for use in the ITER vacuum pumping system

    International Nuclear Information System (INIS)

    Kirchhof, U.; Kammerer, B.; Perinic, D.

    1995-04-01

    This report is a description of the design and construction of FORTE (Forepumps Test Facility) which has been built in order to enable testing of the pumping speeds of prototypical mechanical forepumps connected in series, as proposed for the ITER forepump system. Three NORMETEX pumps (1300, 600, 60 m 3 /h) and one METAL BELLOWS pump (6m 3 /h) have been integrated into the test bench. Measurements of the pumping characteristics were performed, both with the single pumps and with trains of series connected pumps, using the gases N 2 , H 2 , D 2 , He as well as ITER typical gas mixture. The results of the tests are presented. (orig.)

  20. Weld distortion prediction and control of the ITER vacuum vessel manufacturing mock-ups

    International Nuclear Information System (INIS)

    Ottolini, Marco; Barbensi, Andrea

    2014-01-01

    The fabrication of the ITER Vacuum Vessel Sectors is an unprecedented challenge, due to their dimensions, the close tolerances, the complex 'D' shape. The technological issues were faced by the production of full scale mock ups to confirm the manufacturing feasibility to achieve very tight tolerances and qualify the main manufacturing processes, by a step by step welding distortion control, by the qualification of not conventional NDT inspection techniques and by innovative 3D dimensional inspections. The Supplier is required to fabricate at least two mock ups, inboard and outboard, related to the manufacturing method of the VV Sectors, to demonstrate the control of the welding distortions to achieve tolerances, optimizing welding sequences and calibrating of welding distortions computer simulations. The stages of this preparatory activity are: prediction of welding distortion for fabrication mock ups representative of selected segments; demonstration that distortion predictions are consistent with experimental results from 3D dimensional inspection; understanding of reasons of possible deviations between numerical and experimental results and definition of action to solve these issues; demonstration that possible calculation simplifications, adopted to speed up the analysis process, do not affect significantly the welding distortion prediction. This paper describes the weld distortion prediction and control on the manufacturing mock-ups of ITER Vacuum Vessel Sectors, with particular emphasis to the lessons learned. (authors)

  1. Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

    International Nuclear Information System (INIS)

    Di Gironimo, G.; Carfora, D.; Esposito, G.; Lanzotti, A.; Marzullo, D.; Siuko, M.

    2015-01-01

    Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed

  2. Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

    Energy Technology Data Exchange (ETDEWEB)

    Di Gironimo, G., E-mail: giuseppe.digironimo@unina.it [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Carfora, D. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland); Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Esposito, G.; Lanzotti, A.; Marzullo, D. [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Siuko, M. [VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland)

    2015-05-15

    Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed.

  3. Test facility TIMO for testing the ITER model cryopump

    International Nuclear Information System (INIS)

    Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

    2001-01-01

    Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L' Air Liquide in the form of a NET contract. (author)

  4. Test facility TIMO for testing the ITER model cryopump

    International Nuclear Information System (INIS)

    Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

    1999-01-01

    Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L'Air Liquide in the form of a NET contract. (author)

  5. Chatter suppression methods of a robot machine for ITER vacuum vessel assembly and maintenance

    International Nuclear Information System (INIS)

    Wu, Huapeng; Wang, Yongbo; Li, Ming; Al-Saedi, Mazin; Handroos, Heikki

    2014-01-01

    Highlights: •A redundant 10-DOF serial-parallel hybrid robot for ITER assembly and maintains is presented. •A dynamic model of the robot is developed. •A feedback and feedforward controller is presented to suppress machining vibration of the robot. -- Abstract: In the process of assembly and maintenance of ITER vacuum vessel (ITER VV), various machining tasks including threading, milling, welding-defects cutting and flexible hose boring are required to be performed from inside of ITER VV by on-site machining tools. Robot machine is a promising option for these tasks, but great chatter (machine vibration) would happen in the machining process. The chatter vibration will deteriorate the robot accuracy and surface quality, and even cause some damages on the end-effector tools and the robot structure itself. This paper introduces two vibration control methods, one is passive and another is active vibration control. For the passive vibration control, a parallel mechanism is presented to increase the stiffness of robot machine; for the active vibration control, a hybrid control method combining feedforward controller and nonlinear feedback controller is introduced for chatter suppression. A dynamic model and its chatter vibration phenomena of a hybrid robot is demonstrated. Simulation results are given based on the proposed hybrid robot machine which is developed for the ITER VV assembly and maintenance

  6. Chatter suppression methods of a robot machine for ITER vacuum vessel assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Huapeng; Wang, Yongbo, E-mail: yongbo.wang@lut.fi; Li, Ming; Al-Saedi, Mazin; Handroos, Heikki

    2014-10-15

    Highlights: •A redundant 10-DOF serial-parallel hybrid robot for ITER assembly and maintains is presented. •A dynamic model of the robot is developed. •A feedback and feedforward controller is presented to suppress machining vibration of the robot. -- Abstract: In the process of assembly and maintenance of ITER vacuum vessel (ITER VV), various machining tasks including threading, milling, welding-defects cutting and flexible hose boring are required to be performed from inside of ITER VV by on-site machining tools. Robot machine is a promising option for these tasks, but great chatter (machine vibration) would happen in the machining process. The chatter vibration will deteriorate the robot accuracy and surface quality, and even cause some damages on the end-effector tools and the robot structure itself. This paper introduces two vibration control methods, one is passive and another is active vibration control. For the passive vibration control, a parallel mechanism is presented to increase the stiffness of robot machine; for the active vibration control, a hybrid control method combining feedforward controller and nonlinear feedback controller is introduced for chatter suppression. A dynamic model and its chatter vibration phenomena of a hybrid robot is demonstrated. Simulation results are given based on the proposed hybrid robot machine which is developed for the ITER VV assembly and maintenance.

  7. Multi-scenario evaluation and specification of electromagnetic loads on ITER vacuum vessel

    International Nuclear Information System (INIS)

    Rozov, Vladimir; Martinez, J.-M.; Portafaix, C.; Sannazzaro, G.

    2014-01-01

    Highlights: • We present the results of multi-scenario analysis of EM loads on ITER vacuum vessel (VV). • The differentiation of models provides the economic way to perform big amount of calculations. • Functional approximation is proposed for distributed data/FE/numerical results specification. • Examples of specification of the load profiles by trigonometric polynomials (DHT) are given. • Principles of accounting for toroidal asymmetry at EM interactions in tokamak are considered. - Abstract: The electro-magnetic (EM) transients cause mechanical forces, which represent one of the most critical loads for the ITER vacuum vessel (VV). The paper is focused on the results of multi-scenario analysis and systematization of these EM loads, including specifically addressed pressures on shells and the net vertical force. The proposed mathematical model and computational technology, based on the use of integral parameters and operational analysis methods, enabled qualitative and quantitative analysis of the problem, time-efficient computations and systematic assessment of a large number of scenarios. The obtained estimates, found envelopes and peak values exemplify the principal loads on the VV and provide a database to support engineering load specifications. Special attention is given to the challenge of specification and documenting of the results in a form, suitable for using the data in engineering applications. The practical aspects of specification of distributed data, such as experimental and finite-element (FE) results, by analytical interpolants are discussed. The example of functional approximation of the load profiles by trigonometric polynomials based on discrete Hartley transform (DHT) is given

  8. Multi-scenario evaluation and specification of electromagnetic loads on ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Rozov, Vladimir, E-mail: vladimir.rozov@iter.org; Martinez, J.-M.; Portafaix, C.; Sannazzaro, G.

    2014-10-15

    Highlights: • We present the results of multi-scenario analysis of EM loads on ITER vacuum vessel (VV). • The differentiation of models provides the economic way to perform big amount of calculations. • Functional approximation is proposed for distributed data/FE/numerical results specification. • Examples of specification of the load profiles by trigonometric polynomials (DHT) are given. • Principles of accounting for toroidal asymmetry at EM interactions in tokamak are considered. - Abstract: The electro-magnetic (EM) transients cause mechanical forces, which represent one of the most critical loads for the ITER vacuum vessel (VV). The paper is focused on the results of multi-scenario analysis and systematization of these EM loads, including specifically addressed pressures on shells and the net vertical force. The proposed mathematical model and computational technology, based on the use of integral parameters and operational analysis methods, enabled qualitative and quantitative analysis of the problem, time-efficient computations and systematic assessment of a large number of scenarios. The obtained estimates, found envelopes and peak values exemplify the principal loads on the VV and provide a database to support engineering load specifications. Special attention is given to the challenge of specification and documenting of the results in a form, suitable for using the data in engineering applications. The practical aspects of specification of distributed data, such as experimental and finite-element (FE) results, by analytical interpolants are discussed. The example of functional approximation of the load profiles by trigonometric polynomials based on discrete Hartley transform (DHT) is given.

  9. Results from ITER Vacuum Vessel Sector Manufacturing Development in Europe

    International Nuclear Information System (INIS)

    Jones, L.

    2006-01-01

    Significant results have been achieved since the previous SOFT conference, when the manufacturing development work required to prepare for the ITER Vacuum Vessel Sector was described. The contract for the manufacture of a full-size, 20 Ton poloidal part of the inboard section, fabricated according to the ITER reference manufacturing route, including bracing fixtures, welding applications, restraint effects, and fit-up aspects is approaching completion. Since the main aim of the work is to establish the practicability of achieving the tight dimensional tolerances, an accompanying SYSWELD analysis programme has been validation by instrumented welding coupons, and then used for predicting the distortion of the actual construction. A local machining tool has been developed to allow the requirement for machining of the cylindrical features at a late stage of manufacture. Experimental and analytical work has also been carried out to establish the possibility of 3-D cold-forming large sections of walls of the VV. A manufacturing programme to validate an alternative method of fabricating parts of the double-walled VV, utilising e-beam welding only and avoiding the quality issues of the one-sided access and inspection of the closing welds is presented. This paper describes the results of the manufacturing development programme and the future activities. (author)

  10. Bolted Ribs Analysis for the ITER Vacuum Vessel using Finite Element Submodelling Techniques

    Energy Technology Data Exchange (ETDEWEB)

    Zarzalejos, José María, E-mail: jose.zarzalejos@ext.f4e.europa.eu [External at F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Fernández, Elena; Caixas, Joan; Bayón, Angel [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Polo, Joaquín [Iberdrola Ingeniería y Construcción, Avenida de Manoteras 20, 28050 Madrid (Spain); Guirao, Julio [Numerical Analysis Technologies, S L., Marqués de San Esteban 52, Entlo, 33209 Gijon (Spain); García Cid, Javier [Iberdrola Ingeniería y Construcción, Avenida de Manoteras 20, 28050 Madrid (Spain); Rodríguez, Eduardo [Mechanical Engineering Department EPSIG, University of Oviedo, Gijon (Spain)

    2014-10-15

    Highlights: • The ITER Vacuum Vessel Bolted Ribs assemblies are modelled using Finite Elements. • Finite Element submodelling techniques are used. • Stress results are obtained for all the assemblies and a post-processing is performed. • All the elements of the assemblies are compliant with the regulatory provisions. • Submodelling is a time-efficient solution to verify the structural integrity of this type of structures. - Abstract: The ITER Vacuum Vessel (VV) primary function is to enclose the plasmas produced by the ITER Tokamak. Since it acts as the first radiological barrier of the plasma, it is classified as a class 2 welded box structure, according to RCC-MR 2007. The VV is made of an inner and an outer D-shape, 60 mm-thick double shell connected through thick massive bars (housings) and toroidal and poloidal structural stiffening ribs. In order to provide neutronic shielding to the ex-vessel components, the space between shells is filled with borated steel plates, called In-Wall Shielding (IWS) blocks, and water. In general, these blocks are connected to the IWS ribs which are connected to adjacent housings. The development of a Finite Element model of the ITER VV including all its components in detail is unaffordable from the computational point of view due to the large number of degrees of freedom it would require. This limitation can be overcome by using submodelling techniques to simulate the behaviour of the bolted ribs assemblies. Submodelling is a Finite Element technique which allows getting more accurate results in a given region of a coarse model by generating an independent, finer model of the region under study. In this paper, the methodology and several simulations of the VV bolted ribs assemblies using submodelling techniques are presented. A stress assessment has been performed for the elements involved in the assembly considering possible types of failure and including stress classification and categorization techniques to analyse

  11. Detailed Design and Fabrication Method of the ITER Vacuum Vessel Ports

    International Nuclear Information System (INIS)

    Hee-Jae Ahn; Kwon, T.H.; Hong, Y.S.

    2006-01-01

    The engineering design of the ITER vacuum vessel (VV) has been progressed by the ITER International Team (IT) with the cooperation of several participant teams (PT). The VV and ports are the components allocated to Korea for the construction of the ITER. Hyundai Heavy Industries has been involved in the structural analysis, detailed design and development of the fabrication method of the upper and lower ports within the framework of the ITER transitional arrangements (ITA). The design of the port structures has been investigated to validate and to improve the conceptual designs of the ITER IT and other PT. The special emphasis was laid on the flange joint between the port extension and the in-port plug to develop the design of the upper port. The modified design with a pure friction type flange with forty-eight pieces of bolts instead of the tangential key is recommended. Furthermore, the alternative flange designs developed by the ITER IT have been analyzed in detail to simplify the lip seal maintenance into the port flange. The structural analyses of the lower RH port have been also performed to verify the capacity for supporting the VV. The maximum stress exceeds the allowable value at the reinforcing block and basement. More elaborate local models have been developed to mitigate the stress concentration and to modify the component design. The fabrication method and the sequence of the detailed fabrication for the ports are developed focusing on the cost reduction as well as the simplification. A typical port structure includes a port stub, a stub extension and a port extension with a connecting duct. The fabrication sequence consists of surface treatment, cutting, forming, cleaning, welding, machining, and non-destructive inspection and test. Tolerance study has been performed to avoid the mismatch of each fabricated component and to obtain the suitable tolerances in the assembly at the shop and site. This study is based on the experience in the fabrication of

  12. International Thermonuclear Experimental Reactor (ITER) neutral beam design

    International Nuclear Information System (INIS)

    Myers, T.J.; Brook, J.W.; Spampinato, P.T.; Mueller, J.P.; Luzzi, T.E.; Sedgley, D.W.

    1990-10-01

    This report discusses the following topics on ITER neutral beam design: ion dump; neutralizer and module gas flow analysis; vacuum system; cryogenic system; maintainability; power distribution; and system cost

  13. Methodology for dimensional variation analysis of ITER integrated systems

    International Nuclear Information System (INIS)

    Fuentes, F. Javier; Trouvé, Vincent; Cordier, Jean-Jacques; Reich, Jens

    2016-01-01

    Highlights: • Tokamak dimensional management methodology, based on 3D variation analysis, is presented. • Dimensional Variation Model implementation workflow is described. • Methodology phases are described in detail. The application of this methodology to the tolerance analysis of ITER Vacuum Vessel is presented. • Dimensional studies are a valuable tool for the assessment of Tokamak PCR (Project Change Requests), DR (Deviation Requests) and NCR (Non-Conformance Reports). - Abstract: The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on

  14. Methodology for dimensional variation analysis of ITER integrated systems

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes, F. Javier, E-mail: FranciscoJavier.Fuentes@iter.org [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France); Trouvé, Vincent [Assystem Engineering & Operation Services, rue J-M Jacquard CS 60117, 84120 Pertuis (France); Cordier, Jean-Jacques; Reich, Jens [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France)

    2016-11-01

    Highlights: • Tokamak dimensional management methodology, based on 3D variation analysis, is presented. • Dimensional Variation Model implementation workflow is described. • Methodology phases are described in detail. The application of this methodology to the tolerance analysis of ITER Vacuum Vessel is presented. • Dimensional studies are a valuable tool for the assessment of Tokamak PCR (Project Change Requests), DR (Deviation Requests) and NCR (Non-Conformance Reports). - Abstract: The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on

  15. Novel Robot Solutions for Carrying out Field Joint Welding and Machining in the Assembly of the Vacuum Vessel of ITER

    International Nuclear Information System (INIS)

    Pessi, P.

    2009-01-01

    It is necessary to use highly specialized robots in ITER (International Thermonuclear Experimental Reactor) both in the manufacturing and maintenance of the reactor due to a demanding environment. The sectors of the ITER vacuum vessel (VV) require more stringent tolerances than normally expected for the size of the structure involved. VV consists of nine sectors that are to be welded together. The vacuum vessel has a toroidal chamber structure. The task of the designed robot is to carry the welding apparatus along a path with a stringent tolerance during the assembly operation. In addition to the initial vacuum vessel assembly, after a limited running period, sectors need to be replaced for repair. Mechanisms with closed-loop kinematic chains are used in the design of robots in this work. One version is a purely parallel manipulator and another is a hybrid manipulator where the parallel and serial structures are combined. Traditional industrial robots that generally have the links actuated in series are inherently not very rigid and have poor dynamic performance in high speed and high dynamic loading conditions. Compared with open chain manipulators, parallel manipulators have high stiffness, high accuracy and a high force/torque capacity in a reduced workspace. Parallel manipulators have a mechanical architecture where all of the links are connected to the base and to the end-effector of the robot. The purpose of this thesis is to develop special parallel robots for the assembly, machining and repairing of the VV of the ITER. The process of the assembly and machining of the vacuum vessel needs a special robot. By studying the structure of the vacuum vessel, two novel parallel robots were designed and built; they have six and ten degrees of freedom driven by hydraulic cylinders and electrical servo motors. Kinematic models for the proposed robots were defined and two prototypes built. Experiments for machine cutting and laser welding with the 6-DOF robot were

  16. Novel Robot Solutions for Carrying out Field Joint Welding and Machining in the Assembly of the Vacuum Vessel of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pessi, P.

    2009-07-01

    It is necessary to use highly specialized robots in ITER (International Thermonuclear Experimental Reactor) both in the manufacturing and maintenance of the reactor due to a demanding environment. The sectors of the ITER vacuum vessel (VV) require more stringent tolerances than normally expected for the size of the structure involved. VV consists of nine sectors that are to be welded together. The vacuum vessel has a toroidal chamber structure. The task of the designed robot is to carry the welding apparatus along a path with a stringent tolerance during the assembly operation. In addition to the initial vacuum vessel assembly, after a limited running period, sectors need to be replaced for repair. Mechanisms with closed-loop kinematic chains are used in the design of robots in this work. One version is a purely parallel manipulator and another is a hybrid manipulator where the parallel and serial structures are combined. Traditional industrial robots that generally have the links actuated in series are inherently not very rigid and have poor dynamic performance in high speed and high dynamic loading conditions. Compared with open chain manipulators, parallel manipulators have high stiffness, high accuracy and a high force/torque capacity in a reduced workspace. Parallel manipulators have a mechanical architecture where all of the links are connected to the base and to the end-effector of the robot. The purpose of this thesis is to develop special parallel robots for the assembly, machining and repairing of the VV of the ITER. The process of the assembly and machining of the vacuum vessel needs a special robot. By studying the structure of the vacuum vessel, two novel parallel robots were designed and built; they have six and ten degrees of freedom driven by hydraulic cylinders and electrical servo motors. Kinematic models for the proposed robots were defined and two prototypes built. Experiments for machine cutting and laser welding with the 6-DOF robot were

  17. Vacuum system for ISABELLE

    International Nuclear Information System (INIS)

    Hobson, J.P.

    1975-01-01

    An analysis is presented of the proposed vacuum system for the planned ISABELLE storage rings with respect to acceptability and practicality from the vacuum viewport. A comparison is made between the proposed vacuum system and the vacuum system at the CERN ISR, and some comments on various design and operational parameters are made

  18. Fusion Power measurement at ITER

    Energy Technology Data Exchange (ETDEWEB)

    Bertalot, L.; Barnsley, R.; Krasilnikov, V.; Stott, P.; Suarez, A.; Vayakis, G.; Walsh, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-07-01

    Nuclear fusion research aims to provide energy for the future in a sustainable way and the ITER project scope is to demonstrate the feasibility of nuclear fusion energy. ITER is a nuclear experimental reactor based on a large scale fusion plasma (tokamak type) device generating Deuterium - Tritium (DT) fusion reactions with emission of 14 MeV neutrons producing up to 700 MW fusion power. The measurement of fusion power, i.e. total neutron emissivity, will play an important role for achieving ITER goals, in particular the fusion gain factor Q related to the reactor performance. Particular attention is given also to the development of the neutron calibration strategy whose main scope is to achieve the required accuracy of 10% for the measurement of fusion power. Neutron Flux Monitors located in diagnostic ports and inside the vacuum vessel will measure ITER total neutron emissivity, expected to range from 1014 n/s in Deuterium - Deuterium (DD) plasmas up to almost 10{sup 21} n/s in DT plasmas. The neutron detection systems as well all other ITER diagnostics have to withstand high nuclear radiation and electromagnetic fields as well ultrahigh vacuum and thermal loads. (authors)

  19. 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

  20. Parametric assessments on hydrogenic species transport in CVD-diamond vacuum windows used in ITER ECRH

    Energy Technology Data Exchange (ETDEWEB)

    Moreno, C.; Sedano, L.A.; Fernandez, A. [EURATOM-CIEMAT Association, Madrid (Spain)

    2007-07-01

    Insulators used as H and CD and Diagnostic vacuum windows (VW) in ITER may become modified by surface intake and bulk transport of hydrogenic species. VW, operating under severe radiation levels, have a primary safety role as tritium confinement barriers. Ionizing radiation enhances the (H') uptake and release at surfaces and diffusion rates in the bulk. Radiation damage modifies the material's bulk trapped inventories by increasing steady state trapping centre concentrations. An experimental programme is ongoing at CIEMAT, to quantify radiation effects on H transport characteristics and also the possible impact on the VW. The reference material for ECRH VW is CVD diamond. As a parallel activity, parametric transport assessments are being made in order to obtain a wide evaluation of permeation fluxes, ranges, and soluted/trapped inventories in CVD diamond. Transport models have been developed based on extended capabilities of finite differences integrator tool TMAP7. Special attention is paid to radiation parameters defining inputs acting on transport magnitudes. These inputs have been analysed by using ionizing/damage radiation transport tools such as MCNPX/SRIM. VW operational scenarios are discussed with special attention being paid to the ITER design assumptions for the values of H-species source terms (neutrals and implanted) in the ECRH system. The available material transport database with and without radiation is discussed and taken as reference for this parametric exercise. Permeation fluxes through base materials are shown to be below DRG limits established for ITER. (orig.)

  1. Development and control towards a parallel water hydraulic weld/cut robot for machining processes in ITER vacuum vessel

    International Nuclear Information System (INIS)

    Wu Huapeng; Handroos, Heikki; Pessi, Pekka; Kilkki, Juha; Jones, Lawrence

    2005-01-01

    This paper presents a special robot, able to carry out welding and machining processes from inside the ITER vacuum vessel (VV), consisting of a five degree-of-freedom parallel mechanism, mounted on a carriage driven by two electric motors on a rack. The kinematic design of the robot has been optimised for ITER access and a hydraulically actuated pre-prototype built. A hybrid controller is designed for the robot, including position, speed and pressure feedback loops to achieve high accuracy and high dynamic performances. Finally, the experimental tests are given and discussed

  2. Structural Analysis for an Upper Port of the ITER Vacuum Vessel

    International Nuclear Information System (INIS)

    Yun-Seok Hong; Kwon, T. K.; Ahn, H. J.; Kim, Y.K.; Lee, C.D.

    2006-01-01

    The ITER vacuum vessel (VV) has numerous openings for the port structures including upper, equatorial, and lower ports used for equipment installation, utility feed through, vacuum pumping, and access into the vessel for maintenance. Every upper port, slanted upward slightly, has a trapezoidal/rectangular cross-section and consists of a port stub, a stub extension and a port extension with a connecting duct. To investigate the structural integrity and to increase the structural reliability of the VV and ports, the structural analyses of the upper port structure have been performed. The global structural analysis of the upper port with the in-port components has been carried out. The local analyses of a tangential key, an upper port flange, a connecting duct and a sealing unit have been performed. The design loads are dead weight, normal and abnormal pressure load, electromagnetic load, and seismic load in consideration of the dynamic amplification factors. The stress analyses were performed in a nonlinear elastic approach taking into account the contact surface between port extension flange and port plug flange. Two advanced designs from the ITER international team have been reviewed. To verify the strength of the reinforcing ribs for the connecting duct and of the fastening/sealing units, the local analyses utilizing the sub-modeling technique have been performed. The ASME code and the ITER design criteria were applied for the evaluation of the structural analysis results from the global and local analyses. The clearance between a port and a plug to accommodate the plug deformation has been assessed. The upper port flange based on the original design could withstand design loads, but there could be a gap on the flange surface under the design condition. The modified flange design, which is under the bolt friction only without tangential key was proposed. The deflection of the plug for an advanced design with a removable flange is higher than that for the original

  3. Design and R and D for the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Ioki, K.; Johnson, G.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Iizuka, T.; Parker, R.; Koizumi, K.; Kuzmin, E.; Maisonnier, D.; Nelson, B.

    1998-01-01

    The current design and key R and D results for the Vacuum Vessel (VV) for the International Thermonuclear Experimental Reactor (ITER) are presented. During the past two years the basic VV design has remained unchanged. Additional details have been defined in key areas and recent R and D results have indicated where further improvements can be made. R and D results have also confirmed the feasibility of important aspects of the design such as limiting weld distortions to acceptable levels and achieving required tolerances with a large welded structure. Recent design progress includes the development of a structural design strategy for the VV, modification of the inboard structure, employment of ferromagnetic material between the VV shells, and confirmation of the cooling characteristics for the VV. This report presents the current design and how it has been affected by R and D results. (authors)

  4. Design and R and D for the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K.; Johnson, G.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Iizuka, T.; Parker, R. [ITER Joint Work Site, Garching (Germany); Koizumi, K. [Japan Atomic Energy Research Inst., Naka (Japan); Kuzmin, E. [Efremov Insitute, Saint Petersburg (Russian Federation); Maisonnier, D. [NET Team, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    The current design and key R and D results for the Vacuum Vessel (VV) for the International Thermonuclear Experimental Reactor (ITER) are presented. During the past two years the basic VV design has remained unchanged. Additional details have been defined in key areas and recent R and D results have indicated where further improvements can be made. R and D results have also confirmed the feasibility of important aspects of the design such as limiting weld distortions to acceptable levels and achieving required tolerances with a large welded structure. Recent design progress includes the development of a structural design strategy for the VV, modification of the inboard structure, employment of ferromagnetic material between the VV shells, and confirmation of the cooling characteristics for the VV. This report presents the current design and how it has been affected by R and D results. (authors)

  5. Cold Vacuum Drying (CVD) Facility Vacuum Purge System Chilled Water System Design Description. System 47-4

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    2000-01-01

    This system design description (SDD) addresses the Vacuum Purge System Chilled Water (VPSCHW) system. The discussion that follows is limited to the VPSCHW system and its interfaces with associated systems. The reader's attention is directed to Drawings H-1-82162, Cold Vacuum Drying Facility Process Equipment Skid PandID Vacuum System, and H-1-82224, Cold Vacuum Drying Facility Mechanical Utilities Process Chilled Water PandID. Figure 1-1 shows the location and equipment arrangement for the VPSCHW system. The VPSCHW system provides chilled water to the Vacuum Purge System (VPS). The chilled water provides the ability to condense water from the multi-canister overpack (MCO) outlet gases during the MCO vacuum and purge cycles. By condensing water from the MCO purge gas, the VPS can assist in drying the contents of the MCO

  6. Engineering analysis of ITER In-Vessel Viewing System guide tube

    Energy Technology Data Exchange (ETDEWEB)

    Casal, Natalia, E-mail: natalia.casal@iter.org [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Bates, Philip [Fusion for Energy, Barcelona (Spain); Bede, Ottó [Oxford Technologies Ltd., Abingdon (United Kingdom); Damiani, Carlo; Dubus, Gregory [Fusion for Energy, Barcelona (Spain); Omran, Hassan [Oxford Technologies Ltd., Abingdon (United Kingdom); Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Puiu, Adrian [Fusion for Energy, Barcelona (Spain); Reichle, Roger; Suárez, Alejandro; Walker, Christopher; Walsh, Michael [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France)

    2015-10-15

    Highlights: • Conceptual design of IVVS Loads action on IVVS Dominant loads. • Seismic and baking conditions. • No active cooling needed for IVVS. • IVVS requires additional support points to avoid excessive deformation. - Abstract: The In Vessel Viewing System (IVVS) will be one of the essential machine diagnostic systems at ITER to provide information about the status of in-vessel and plasma facing components and to evaluate the dust inside the Vacuum Vessel. The current design consists of six scanning probes and their deployment systems, which are placed in dedicated ports at the divertor level. These units are located in resident guiding tubes 10 m long, which allow the IVVS probes to go from their storage location to the scanning position by means of a simple straight translation. Moreover, each resident tube is supported inside the corresponding Vacuum Vessel and Cryostat port extensions, which are part of the primary confinement barrier. As the Vacuum Vessel and the Cryostat will move with respect to each other during operation (especially during baking) and during incidents and accidents (disruptions, vertical displacement events, seismic events), the structural integrity of the resident tube and the surrounding vacuum boundaries would be compromised if the required flexibility and supports are not appropriately assured. This paper focuses on the integration of the present design of the IVVS into the Vacuum Vessel and Cryostat environment. It presents the adopted strategy to withstand all the main interfacing loads without damaging the confinement barriers and the corresponding analysis supporting it.

  7. Assessment of the dynamic behaviours of the ITER Vacuum Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Blocki, J., E-mail: jacek.blocki@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Combescure, D. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Mazzone, G. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► The cyclic symmetry structure with special boundary conditions has been analyzed. ► Results based on the FE solid model and on the FE shell model have been compared. ► The effect of the missing mass contained has been checked. -- Abstract: The dynamic behaviour of the ITER Vacuum Vessel (VV) under seismic loads will be assessed by carrying out the modal analysis and then by applying the response spectrum method which describes earthquake motions. The effect of the missing mass is included in this last analysis. Numerical results are based on two different Finite Element (FE) models and on three different methods by which natural frequencies and mode shapes are defined. It means, it is applied the cyclic symmetry method, the component mode synthesis method and the 360° FE model of the VV. Comparisons between obtained results for the different models and methods are presented.

  8. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    International Nuclear Information System (INIS)

    Nakahira, M.; Takeda, N.; Kakudate, S.; Onozuka, M.

    2008-01-01

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV

  9. Design and structural analysis of support structure for ITER vacuum vessel

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Ohmori, Junji; Nakahira, Masataka; Shibanuma, Kiyoshi

    2004-01-01

    The International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed as a new concept, which is deferent from the current design, i.e., the VV support is directly connected to the toroidal coils (TF coils). This independent concept has two advantages comparing to the current one: (1) thermal load due to the temperature deference between VV and TF coils becomes lower and (2) the TF coils are categorized as non-safety components because of its independence from VV. Stress Analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coils is found to be 15 mm, much less than the current design clearance of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support. (author)

  10. The radiation analyses of ITER lower ports

    International Nuclear Information System (INIS)

    Petrizzi, L.; Brolatti, G.; Martin, A.; Loughlin, M.; Moro, F.; Villari, R.

    2010-01-01

    The ITER Vacuum Vessel has upper, equatorial, and lower ports used for equipment installation, diagnostics, heating and current drive systems, cryo-vacuum pumping, and access inside the vessel for maintenance. At the level of the divertor, the nine lower ports for remote handling, cryo-vacuum pumping and diagnostic are inclined downwards and toroidally located each every 40 o . The cryopump port has additionally a branch to allocate a second cryopump. The ports, as openings in the Vacuum Vessel, permit radiation streaming out of the vessel which affects the heating in the components in the outer regions of the machine inside and outside the ports. Safety concerns are also raised with respect to the dose after shutdown at the cryostat behind the ports: in such zones the radiation dose level must be kept below the regulatory limit to allow personnel access for maintenance purposes. Neutronic analyses have been required to qualify the ITER project related to the lower ports. A 3-D model was used to take into account full details of the ports and the lower machine surroundings. MCNP version 5 1.40 has been used with the FENDL 2.1 nuclear data library. The ITER 40 o model distributed by the ITER Organization was developed in the lower part to include the relevant details. The results of a first analysis, focused on cryopump system only, were recently published. In this paper more complete data on the cryopump port and analysis for the remote handling port and the diagnostic rack are presented; the results of both analyses give a complete map of the radiation loads in the outer divertor ports. Nuclear heating, dpa, tritium production, and dose rates after shutdown are provided and the implications for the design are discussed.

  11. Analysis of the steady state hydraulic behaviour of the ITER blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A.; Vallone, E. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • Nominal steady state hydraulic behaviour of ITER blanket standard sector cooling system has been investigated. • Numerical simulations have been run adopting a qualified thermal-hydraulic system code. • Hydraulic characteristic functions and coolant mass flow rates, velocities and pressure drops have been assessed. • Most of the considered circuits are able to effectively cool blanket modules, meeting ITER requirements. - Abstract: The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised.

  12. 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

  13. 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

  14. ITER task D316 (1996): design review of isotope separation system (WBS 3.2 B) and water detritiation system (WBS 3.2 E)

    International Nuclear Information System (INIS)

    Sood, S.K.; Fong, C.

    1997-05-01

    The design review performed on the ITER Isotope Separation System and the Water Detritiation System are summarized. The objectives of the task are: to produce a Design Description Document for the Feed Treatment and Vacuum system for the Water Detritiation system; to review the process system operation and control philosophy for the Water Detritiation System; to review the equipment arrangement drawings where available. 1 fig., 3 refs

  15. Status of the EU domestic agency electromagnetic analyses of ITER vacuum vessel and blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Testoni, P., E-mail: pietro.testoni@f4e.europa.eu [Fusion for Energy, Josep Plá n. 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Albanese, R. [Association Euratom/ENEA/CREATE, DIEL, Università Federico II di Napoli, Napoli 80125 (Italy); Lucca, F.; Roccella, M. [L.T. Calcoli S.a.S. Piazza Prinetti, 26/B, Merate, Lecco (Italy); Portone, A. [Fusion for Energy, Josep Plá n. 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Rubinacci, G. [Association Euratom/ENEA/CREATE, DIEL, Università Federico II di Napoli, Napoli 80125 (Italy); Ventre, S.; Villone, F. [Association Euratom/ENEA/CREATE, DAEIMI, Università di Cassino, Cassino 03043 (Italy)

    2013-10-15

    Highlights: Eddy and halo currents and corresponding Lorentz forces on the ITER vacuum vessel and blanket modules have been computed. VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge have been simulated. The maximum vertical force in the VV (about 120 MN downwards) is experienced in VDE-DW-SLOW cat III. For the FW panel of blanket 18 the most demanding load case is the VDE downward cat III producing a radial torque of about 110 kNm. For the FW of blanket module 10 the most demanding load case is the VDE upward exp cat III producing a poloidal torque of about 130 kNm. -- Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained.

  16. Status of the EU domestic agency electromagnetic analyses of ITER vacuum vessel and blanket modules

    International Nuclear Information System (INIS)

    Testoni, P.; Albanese, R.; Lucca, F.; Roccella, M.; Portone, A.; Rubinacci, G.; Ventre, S.; Villone, F.

    2013-01-01

    Highlights: Eddy and halo currents and corresponding Lorentz forces on the ITER vacuum vessel and blanket modules have been computed. VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge have been simulated. The maximum vertical force in the VV (about 120 MN downwards) is experienced in VDE-DW-SLOW cat III. For the FW panel of blanket 18 the most demanding load case is the VDE downward cat III producing a radial torque of about 110 kNm. For the FW of blanket module 10 the most demanding load case is the VDE upward exp cat III producing a poloidal torque of about 130 kNm. -- Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained

  17. Primary design and operation analysis of ITER air transfer system

    International Nuclear Information System (INIS)

    Wang Haitian; Li Ge; Qin Shijun

    2010-01-01

    Air transfer system (ATS) is a remote handling transfer, which can work in the nuclear radiation environment and can be driven by the electricity fully. Its motion power is provided by several servo motors. The remote control technology of ATS, which is China taking part in the plan of international Tokamak experimental reactor (ITER) and grasping this technology, is one of key technologies of ITER. The remote handling technology can lay the foundation for developing demonstration nuclear fusion power plant in China on self-reliance. Because there is gamma irradiation and hazard material in these ITER parts, all required maintenance of port plugs and inner components are been transmitted by ATS. The pick-up or drop-off these components are completed by means of a remotely controlled TCS system between the Vacuum Vessel and the Hot Cell through the bridge-gallery. Tokamak building includes three floors, including upper port, equatorial port and lower port, linked by a lift. According to each port level configuration and safety requirement, the radius of curvature with ATS trajectory is optimized, and a trajectory of each level is determined by positioned guidance beacons. At last, the results of computer aided design (CAD) show single trajectory guidance of ATS in each level is available. (authors)

  18. Qualification of phased array ultrasonic examination on T-joint weld of austenitic stainless steel for ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, G.H. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Park, C.K., E-mail: love879@hanmail.net [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Jin, S.W.; Kim, H.S.; Hong, K.H.; Lee, Y.J.; Ahn, H.J.; Chung, W. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Jung, Y.H.; Roh, B.R. [Hyundai Heavy Industries Co. Ltd., Ulsan 682-792 (Korea, Republic of); Sa, J.W.; Choi, C.H. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • PAUT techniques has been developed by Hyundai Heavy Industries Co., LTD (HHI) and Korea Domestic Agency (KODA) to verify and settle down instrument calibration, test procedures, image processing, and so on. As the first step of development for PAUT technique, Several dozens of qualification blocks with artificial defects, which are parallel side drilled hole, embedded lack of fusion, embedded repair weld notch, and so on, have been designed and fabricated to simulate all potential defects during welding process. Real UT qualification group-1 for T-joint weld was successfully conducted in front of ANB inspector. • In this paper, remarkable progresses of UT qualification are presented for ITER vacuum vessel. - Abstract: Full penetration welding and 100% volumetric examination are required for all welds of pressure retaining parts of the ITER Vacuum Vessel (VV) according to RCC-MR Code and French Order of Nuclear Pressure Equipment (ESPN). The NDE requirement is one of important technical issues because radiographic examination (RT) is not applicable to many welding joints. Therefore the ultrasonic examination (UT) has been selected as an alternative method. Generally the UT on the austenitic welds is regarded as a great challenge due to the high attenuation and dispersion of the ultrasonic signal. In this paper, Phased array ultrasonic examination (PAUT) has been introduced on double sided T-shape austenitic welds of the ITER VV as a major NDE method as well as RT. Several dozens of qualification blocks with artificial defects, which are parallel side drilled hole, embedded lack of fusion, embedded repair weld notch, embedded parallel vertical notch, and so on, have been designed and fabricated to simulate all potential defects during welding process. PAUT techniques on the thick austenitic welds have been developed taking into account the acceptance criteria. Test procedure including calibration of equipment is derived and qualified through

  19. Summarized results of the cryosorption panel test programme for the ITER cryopumping system

    International Nuclear Information System (INIS)

    Day, C.; Haas, H.; Mack, A.; Kazakovsky, N.T.; Murdoch, D.K.; Roehrig, D.; Saksagansky, G.L.

    2001-01-01

    A reliable but versatile primary cryopumping system is required for high vacuum pumping of the ITER torus during all phases of plasma operation. To achieve that goal, an extensive R and D programme has been performed within the framework of the Nuclear Fusion Project of FZK, supported by the European Communities under the European Fusion Technology Programme. The present paper covers that part of the programme, which focuses on the pumping speed of the recommended cryopanel type and the various aspects of the charcoal-bonding system in the cryogenic temperature range. It is demonstrated that the investigated cryosorption panels exhibit a very good behaviour with respect to pumping efficiency, long-term thermomechanical endurance and compatibility with tritium. The recommended cryopump design was therefore chosen as point design for ITER. (author)

  20. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, M. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)], E-mail: nakahira.masataka@jaea.go.jp; Takeda, N.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Onozuka, M. [Mitsubishi Nuclear Energy Systems, Inc., 1700K Street NW, Suite 440, Washington, DC 20006 (United States)

    2008-12-15

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV.

  1. 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

  2. Structural analysis of ITER multi-purpose deployer

    International Nuclear Information System (INIS)

    Manuelraj, Manoah Stephen; Dutta, Pramit; Gotewal, Krishan Kumar; Rastogi, Naveen; Tesini, Alessandro; Choi, Chang-Hwan

    2016-01-01

    Highlights: • System modelling for structural analysis of the Multi-Purpose Deployer (MPD). • Finite element modeling of the Multi-Purpose Deployer (MPD). • Static, modal and seismic response analysis of the Multi-Purpose Deployer (MPD). • Iterative structural analysis and design update to satisfy the structural criteria. • Modal analysis for various kinematic configurations. • Reaction force calculations on the interfacing systems. - Abstract: The Multi-Purpose Deployer (MPD) is a general purpose ITER in-vessel remote handling (RH) system. The main handling equipment, known as the MPD Transporter, consists of a series of linked bodies, which provide anchoring to the vacuum vessel port and an articulated multi-degree of freedom motion to perform various in-vessel maintenance tasks. During the in-vessel operations, the structural integrity of the system should be guaranteed against various operational and seismic loads. This paper presents the structural analysis results of the concept design of the MPD Transporter considering the seismic events. Static structural, modal and frequency response spectrum analyses have been performed to verify the structural integrity of the system, and to provide reaction forces to the interfacing systems such as vacuum vessel and cask. Iterative analyses and design updates are carried out based on the reference design of the system to improve the structural behavior of the system. The frequency responses of the system in various kinematics and payloads are assessed.

  3. Structural damages prevention of the ITER vacuum vessel and ports by elasto-plastic analysis with regards to RCC-MR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, Jean-Marc, E-mail: jean-marc.martinez@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Jun, Chang Hoon; Portafaix, Christophe; Alekseev, Alexander; Sborchia, Carlo; Choi, Chang-Ho [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Albin, Vincent [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France); Borrelly, Stephane [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Cambazar, Magali [Assystem EOS, 117 rue Jacquard, 84120 Pertuis (France); Gaucher, Thomas [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France); Sfarni, Samir; Tailhardat, Olivier [Assystem EOS, 117 rue Jacquard, 84120 Pertuis (France)

    2015-10-15

    Highlights: • ITER vacuum vessel (VV) is a part of the first barrier to confine the plasma. • ITER VV as NPE necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, and conformance testing and quality assurance, i.e. ANB. • Several types of damages have to be prevented in order to guarantee the structural integrity with regards to RCC-MR. • It is usual to employ non-linear analysis when the “classical” elastic analysis reaches its limit of linear application. • Several structural analyses were performed with many different global and local models of the whole ITER VV. - Abstract: Several types of damages have to be prevented in order to guarantee the structural integrity of a structure with regards to RCC-MR; the P-type damages which can result from the application to a structure of a steadily and regularly increasing loading or a constant loading and the S-type damages during operational loading conditions which can only result from repeated application of loadings associated to the progressive deformations and fatigue. Following RCC-MR, the S-type damages prevention has to be started only when the structural integrity is guaranteed against P-type damages. The verification of the last one on the ITER vacuum vessel and ports has been performed by limit analysis with elasto-(perfectly)plastic material behavior. It is usual to employ non-linear analysis when the “classical” elastic analysis reaches its limit of linear application. Some elasto-plastic analyses have been performed considering several cyclic loadings to evaluate also more realistic structural margins of the against S-type damages.

  4. Shiva and Argus target diagnostics vacuum systems

    International Nuclear Information System (INIS)

    Glaros, S.S.; Mayo, S.E.; Campbell, D.; Holeman, D.

    1978-09-01

    The normal operation of LLL's Argus and Shiva laser irradiation facilities demand a main vacuum system for the target chamber and a separate local vacuum system for each of the larger appendage dianostics. This paper will describe the Argus and Shiva main vacuum systems, their respective auxiliary vacuum systems and the individual diagnostics with their respective special vacuum requirements and subsequent vacuum systems. Our latest approach to automatic computer-controlled vacuum systems will be presented

  5. ITER diagnostics ex-vessel engineering services

    Energy Technology Data Exchange (ETDEWEB)

    Arumugam, A.P., E-mail: arun.prakash@iter.org; Walker, C.I.; Andrew, P.; Barnsley, R.; Beltran, D.; Bertalot, L.; Dammann, A.; Direz, M.F.; Drevon, J.M.; Encheva, A.; Giacomin, T.; Hourtoule, J.; Kuehn, I.; Lanza, R.; Levesy, B.; Maquet, P.; Patel, K.M.; Patisson, L.; Pitcher, C.S.; Portales, M.; and others

    2013-10-15

    Highlights: • This paper describes about the ITER diagnostics ex-vessel engineering services. • It describes various diagnostics systems, its location and its environment. • Diagnostics interfaces with other services such as the buildings, HVAC, electrical services, cooling water, vacuum, liquid and gas distribution. • All the interfaces with these services are identified and defined. • Buildings services for diagnostics, such as penetrations, local shielding, embedment and temperature control are discussed. -- Abstract: Extensive diagnostics systems will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include measurements of temperature, density, impurity concentration, and particle and energy confinement times. ITER diagnostic systems extend from the center of the Tokamak to the various diagnostic areas, where they are controlled and acquired data is processed. This mainly includes the areas such as ports, port cells, gallery, diagnostics enclosures and cubicle areas. The diagnostics port plugs encloses the front end of the diagnostic systems and the diagnostics building houses the diagnostics equipment, instrumentation and control cubicles. There are several systems providing services to diagnostics. These mainly include ITER buildings, electrical power services, cooling water services, Heating Ventilation and Air Conditioning (HVAC), vacuum services, liquid and gas distribution services, cable engineering, de-tritiation systems, control cubicles, etc. Requirements of these service systems have to be defined, even though many of the diagnostics are at an early stage of development. It is a real challenge to define and to design diagnostics systems considering the constraints imposed by these service systems. This paper summarizes the provision of these services to the individual diagnostics and diagnostics areas

  6. ITER diagnostics ex-vessel engineering services

    International Nuclear Information System (INIS)

    Arumugam, A.P.; Walker, C.I.; Andrew, P.; Barnsley, R.; Beltran, D.; Bertalot, L.; Dammann, A.; Direz, M.F.; Drevon, J.M.; Encheva, A.; Giacomin, T.; Hourtoule, J.; Kuehn, I.; Lanza, R.; Levesy, B.; Maquet, P.; Patel, K.M.; Patisson, L.; Pitcher, C.S.; Portales, M.

    2013-01-01

    Highlights: • This paper describes about the ITER diagnostics ex-vessel engineering services. • It describes various diagnostics systems, its location and its environment. • Diagnostics interfaces with other services such as the buildings, HVAC, electrical services, cooling water, vacuum, liquid and gas distribution. • All the interfaces with these services are identified and defined. • Buildings services for diagnostics, such as penetrations, local shielding, embedment and temperature control are discussed. -- Abstract: Extensive diagnostics systems will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include measurements of temperature, density, impurity concentration, and particle and energy confinement times. ITER diagnostic systems extend from the center of the Tokamak to the various diagnostic areas, where they are controlled and acquired data is processed. This mainly includes the areas such as ports, port cells, gallery, diagnostics enclosures and cubicle areas. The diagnostics port plugs encloses the front end of the diagnostic systems and the diagnostics building houses the diagnostics equipment, instrumentation and control cubicles. There are several systems providing services to diagnostics. These mainly include ITER buildings, electrical power services, cooling water services, Heating Ventilation and Air Conditioning (HVAC), vacuum services, liquid and gas distribution services, cable engineering, de-tritiation systems, control cubicles, etc. Requirements of these service systems have to be defined, even though many of the diagnostics are at an early stage of development. It is a real challenge to define and to design diagnostics systems considering the constraints imposed by these service systems. This paper summarizes the provision of these services to the individual diagnostics and diagnostics areas

  7. ITER EDA Newsletter. V. 3, no. 8

    International Nuclear Information System (INIS)

    1994-08-01

    This ITER EDA (Engineering Design Activities) Newsletter issue reports on the sixth ITER council meeting; introduces the newly appointed ITER director and reports on his address to the ITER council. The vacuum tank for the ITER model coil testing, installed at JAERI, Naka, Japan is also briefly described

  8. Feasibility analysis of vacuum sieve tray for tritium extraction in the HCLL test blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Okino, Fumito, E-mail: fumito.okino@iae.kyoto-u.ac.jp [Kyoto University Institute of Advanced Energy, 611-0011 Gokasho, Uji, Kyoto (Japan); Calderoni, Pattrick [Fusion For Energy, 08019 Barcelona (Spain); Kasada, Ryuta; Konishi, Satoshi [Kyoto University Institute of Advanced Energy, 611-0011 Gokasho, Uji, Kyoto (Japan)

    2016-11-01

    Highlights: • The authors discovered faster mass transport on a droplet falling in a vacuum. • Primary cause of the hydrogen release from droplet is by the oscillation of a droplet. • The spherical oscillation induces the internal advection and enhances mass transfer. • This assumption agreed with previous experimental results. - Abstract: This paper describes the quantitative analysis for the design of a tritium extraction system that uses liquid PbLi droplets in vacuum (Vacuum Sieve Tray, VST), for application to the ITER helium-cooled lithium lead (HCLL) test blanket system (TBS). The parametric dependences of tritium extraction efficiency from the main geometrical features such as initial droplet velocity, nozzle head height, nozzle diameter, and flow rate are discussed. With nozzle diameters between 0.4 and 0.6 mm, extraction efficiency is estimated from 0.77 to 0.96 at the falling height of 0.5 m, with flow rate between 0.2 and 1.0 kg/s. The device has a height of 1.6 m, within the external dimensions of the HCLL Test Blanket Module (TBM), and no additional pumping power is required. The attained results are considered attractive not only for ITER, but also in view of the application of the VST concept as a candidate tritium extraction system for the European Union's demonstration fusion reactor (DEMO). The extraction efficiency of a single droplet column, which is the basis of the design analysis presented, has been validated experimentally with hydrogen. However, further experiments are required on an integrated system with size relevant to the proposed HCLL-TBS design to validate system-level effects, particularly regarding the desorption process in an array of multiple droplets.

  9. Primary design and operation analysis of the ITER air transfer system

    International Nuclear Information System (INIS)

    Wang Haitian; Li Ge; Qin Shijun

    2010-01-01

    Air transfer system (ATS) is a remote handling transfer, which can work in the nuclear radiation environment and can be driven by the electricity fully. Its motion power is provided by several servo motors. The remote control technology of ATS, which is China taking part in the plan of international Tokamak experimental reactor (ITER) and grasping this technology, is one of key technologies of ITER. The remote handling technology can lay the foundation for developing demonstration nuclear fusion power plant in China on self-reliance. Because there is gamma irradiation and hazard material in these ITER parts, all required maintenance of port plugs and inner components are been transmitted by ATS. The pick-up or drop-off these components are completed by means of a remotely controlled TCS system between the Vacuum Vessel and the Hot Cell through the bridge-gallery. Tokamak building includes three floors, including upper port, equatorial port and lower port, linked by a lift. According to each port level configuration and safety requirement, the radius of curvature with ATS trajectory is optimized, and a trajectory of each level is determined by positioned guidance beacons. At last, the results of computer aided design (CAD) show single trajectory guidance of ATS in each level is available. (authors)

  10. Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

    International Nuclear Information System (INIS)

    Onozuka, M.; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H.

    2001-01-01

    Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region

  11. Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H

    2001-11-01

    Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region.

  12. Guidelines for remote handling maintenance of ITER neutral beam line components: Proposal of an alternate supporting system

    International Nuclear Information System (INIS)

    Cordier, J.J.; Bayetti, P.; Hemsworth, R.; David, O.; Friconneau, J.P.

    2007-01-01

    Remote handling (R/H) maintenance of ITER components is one of the main challenges of the ITER project. This type of maintenance shall be operational for the assembly and nuclear phase of exploitation of ITER. It must be considered at a very early stage since it significantly impacts on the components design, interfaces management, assembly, maintenance and integration aspects. A large part of the R/H equipment will be procured by the EU Participating Team, including the whole Neutral Beam R/H Equipment. The Neutral Beam Heating and Current Drive system (NB and CD) design is being revisited by the ITER project. A vertical maintenance scheme is presently considered which may significantly impact on the reference design and associated components and lead to a new design of the NB and CD vacuum tank. In addition, NB line components remote handling solutions are being studied. The neutral beam test facility ITER to be built in Europe in the near future is also based on the vertical NB maintenance scheme of beam line components. New design guidelines compliant for both the ITER NB and CD system and the NB test facility proposed by the CEA association are described in the paper

  13. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 1. Comprehending the vacuum vessel structure

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Nakahira, Masataka

    2006-01-01

    The functions, conditions and structure of vacuum vessel using tokamak fusion machines are explained. The structural standard and code of vacuum vessel, process of vacuum vessel design, and design of ITER vacuum vessel are described. Production and maintenance of ultra high vacuum, confinement of radioactive materials, support of machines in vessel and electromagnetic force, radiation shield, plasma vertical stability, one-turn electric resistance, high temperature baking heat and remove of nuclear heat, reduce of troidal ripple, structural standard, features of safety of nuclear fusion machines, subjects of structural standard of fusion vacuum vessel, design flow of vacuum vessel, establishment of radial build, selections of materials, baking and cooling method, basic structure, structure of special parts, shield structure, and of support structure, and example of design of structure, ITER, are stated. (S.Y.)

  14. Vacuum hot-pressed beryllium and TiC dispersion strengthened tungsten alloy developments for ITER and future fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang, E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Chen, Jiming; Lian, Youyun; Wu, Jihong; Xu, Zengyu; Zhang, Nianman; Wang, Quanming; Duan, Xuro [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Wang, Zhanhong; Zhong, Jinming [Northwest Rare Metal Material Research Institute, CNMC, Ningxia Orient Group Co. Ltd.,No.119 Yejin Road, Shizuishan City, Ningxia,753000 (China)

    2013-11-15

    Beryllium and tungsten have been selected as the plasma facing materials of the ITER first wall (FW) and divertor chamber, respectively. China, as a participant in ITER, will share the manufacturing tasks of ITER first-wall mockups with the European Union and Russia. Therefore ITER-grade beryllium has been developed in China and a kind of vacuum hot-pressed (VHP) beryllium, CN-G01, was characterized for both physical, and thermo-mechanical properties and high heat flux performance, which indicated an equivalent performance to U.S. grade S-65C beryllium, a reference grade beryllium of ITER. Consequently CN-G01 beryllium has been accepted as the armor material of ITER-FW blankets. In addition, a modification of tungsten by TiC dispersion strengthening was investigated and a W–TiC alloy with TiC content of 0.1 wt.% has been developed. Both surface hardness and recrystallization measurements indicate its re-crystallization temperature approximately at 1773 K. Deuterium retention and thermal desorption behaviors of pure tungsten and the TiC alloy were also measured by deuterium ion irradiation of 1.7 keV energy to the fluence of 0.5–5 × 10{sup 18} D/cm{sup 2}; a main desorption peak at around 573 K was found and no significant difference was observed between pure tungsten and the tungsten alloy. Further characterization of the tungsten alloy is in progress.

  15. 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.

  16. ITER plant systems

    International Nuclear Information System (INIS)

    Kolbasov, B.; Barnes, C.; Blevins, J.

    1991-01-01

    As part of a series of documents published by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this publication describes the conceptual design of the ITER plant systems, in particular (i) the heat transport system, (ii) the electrical distribution system, (iii) the requirements for radioactive equipment handling, the hot cell, and waste management, (iv) the supply system for fluids and operational chemicals, (v) the qualitative analyses of failure scenarios and methods of burn stability control and emergency shutdown control, (vi) analyses of tokamak building functions and design requirements, (vii) a plant layout, and (viii) site requirements. Refs, figs and tabs

  17. Technical specification for vacuum systems

    International Nuclear Information System (INIS)

    Khaw, J.

    1987-01-01

    The vacuum systems at the Stanford Linear Accelerator Center (SLAC) are primarily of all-metal construction and operate at pressures from 10 -5 to 10 -11 Torr. The primary gas loads during operation result from thermal desorption and beam-induced desorption from the vacuum chamber walls. These desorption rates can be extremely high in the case of hydrocarbons and other contaminants. These specifications place a major emphasis on eliminating contamination sources. The specifications and procedures have been written to insure the cleanliness and vacuum integrity of all SLAC vacuum systems, and to assist personnel involved with SLAC vacuum systems in choosing and designing components that are compatible with existing systems and meet the quality and reliability of SLAC vacuum standards. The specification includes requirements on design, procurement, fabrication, chemical cleaning, clean room practices, welding and brazing, helium leak testing, residual gas analyzer testing, bakeout, venting, and pumpdown. Also appended are specifications regarding acceptable vendors, isopropyl alcohol, bakeable valve cleaning procedure, mechanical engineering safety inspection, notes on synchrotron radiation, and specifications of numerous individual components

  18. ITER EDA newsletter. V. 5, no. 9

    International Nuclear Information System (INIS)

    1996-09-01

    This issue of the Newsletter on the Engineering Design Activities (EDA) for the ITER project contains an overview of one of the seven large ITER Research and Development Projects identified by the ITER Director, namely the Vacuum Vessel Sector, as well as an account of computer animation created for ITER

  19. Upgrade of DC power supply system in ITER CS model coil test facility

    International Nuclear Information System (INIS)

    Shimono, Mitsugu; Uno, Yasuhiro; Yamazaki, Keita; Kawano, Katsumi; Isono, Takaaki

    2014-03-01

    Objective of the ITER CS Model Coil Test Facility is to evaluate a large scale superconducting conductor for fusion using the Central Solenoid (CS) Model Coil, which can generate a 13T magnetic field in the inner bore with a 1.5 m diameter. The facility is composed of a helium refrigerator / liquefier system, a DC power supply system, a vacuum system and a data acquisition system. The DC power supply system supplies currents to two superconducting coils, the CS Model Coil and an insert coil. A 50-kA DC power supply is installed for the CS Model Coil and two 30 kA DC power supplies are installed for an insert coil. In order to evaluate superconducting performance of a conductor used for ITER Toroidal Field (TF) coils whose operating current is 68 kA, the line for an insert coil is upgraded. A 10 kA DC power supply was added, DC circuit breakers were upgraded, bus bars and current measuring instrument were replaced. In accordance to the upgrade, operation manual was revised. (author)

  20. Hydrogenic Species Transport Assessments in Ceramic Aluminas Used in ITER ICRH H and CD and Diagnostic Systems

    Energy Technology Data Exchange (ETDEWEB)

    Moreno, C.; Sedano, L. A.

    2007-09-27

    Ceramic insulators will be used in the ITER Heating and Current Drive and Diagnostics (H and CD/D) systems as opto-electronic vacuum windows or as feed-troughs. Their performance as materials could come modified by the intake of deuterium-tritium which amounts might be enhanced by ionising radiation effects. Such vacuum windows have a primary safety role as tritium confinement barriers. Tritium transport analyses have major implications on the design and safety assessments of ITER RF H and CD systems. As it is shown, refined tritium transport release-rate models together with detailed parametric studies can precise such assessments. In addition such modeling serves as conceptual framework to quantify precise impact of underlying phenomena (ex. radiation-enhanced diffusion or potential effects of radiation damage on tritium transport through the Vacuum Window) and its fi nal impact on main transport parameters of interest for VW design: permeation flux and D/T inventories. In the present work it has been shown how, for electric implantation of ionized D,T in the VW being the major source for isotopes intake, an hybrid recombination/radiation enhanced diffusion regime determine H-isotopes transport kinetics in the window. Precise values for permeation fluxes and inventories are provided from solution of mass transport equations. Near and medium term work planning is advanced. (Author) 16 refs.

  1. Hydrogenic Species Transport Assessments in Ceramic Aluminas Used in ITER ICRH H and CD and Diagnostic Systems

    International Nuclear Information System (INIS)

    Moreno, C.; Sedano, L. A.

    2007-01-01

    Ceramic insulators will be used in the ITER Heating and Current Drive and Diagnostics (H and CD/D) systems as opto-electronic vacuum windows or as feed-troughs. Their performance as materials could come modified by the intake of deuterium-tritium which amounts might be enhanced by ionising radiation effects. Such vacuum windows have a primary safety role as tritium confinement barriers. Tritium transport analyses have major implications on the design and safety assessments of ITER RF H and CD systems. As it is shown, refined tritium transport release-rate models together with detailed parametric studies can precise such assessments. In addition such modeling serves as conceptual framework to quantify precise impact of underlying phenomena (ex. radiation-enhanced diffusion or potential effects of radiation damage on tritium transport through the Vacuum Window) and its fi nal impact on main transport parameters of interest for VW design: permeation flux and D/T inventories. In the present work it has been shown how, for electric implantation of ionized D,T in the VW being the major source for isotopes intake, an hybrid recombination/radiation enhanced diffusion regime determine H-isotopes transport kinetics in the window. Precise values for permeation fluxes and inventories are provided from solution of mass transport equations. Near and medium term work planning is advanced. (Author) 16 refs

  2. Progress in the integration of Test Blanket Systems in ITER equatorial port cells and in the interfaces definition

    Energy Technology Data Exchange (ETDEWEB)

    Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Beloglazov, S.; Bonagiri, S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Commin, L. [CEA, IRFM, Cadarache (France); Cortes, P.; Giancarli, L.M.; Gliss, C.; Iseli, M.; Lanza, R.; Levesy, B.; Martins, J.-P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Neviere, J.-C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L.; Plutino, D.; Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Swami, H.L. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The design integration of two test blanket systems in ITER port cell is addressed. Black-Right-Pointing-Pointer Definition of interfaces of TBSs with building and other ITER systems is done. Black-Right-Pointing-Pointer Designs of pipe forest, bioshield plug and ancillary equipment unit are described. Black-Right-Pointing-Pointer The maintenance of the two test blanket systems in ITER port cell is considered. Black-Right-Pointing-Pointer The management of the heat and tritium releases in the TBM port cell is described. - Abstract: In the framework of the TBM Program, three ITER vacuum vessel equatorial ports (no. 16, no. 18 and no. 02) have been allocated for the testing of up to six mock-ups of six different DEMO tritium breeding blankets. Each one is called a Test Blanket System (TBS). A TBS consists mainly of the Test Blanket Module (TBM), the in-vessel component facing the plasma, and several ancillary systems, in particular the cooling system and the tritium extraction system. Each port accommodates two TBMs and therefore the two TBSs have to share the corresponding port cell. This paper deals with the design integration aspects of the two TBSs in each port cell performed at ITER Organization (IO) with the corresponding definition of interfaces with other ITER systems. The performed activities have raised several issues that are discussed in the paper and for which design solutions are proposed.

  3. ITER articulated inspection arm (AIA): R and D progress on vacuum and temperature technology for remote handling

    Energy Technology Data Exchange (ETDEWEB)

    Perrot, Y.; Friconneau, J.P. [Robotics and Interactive Systems Unit - CEA/LIST, 92 - Fontenay aux Roses (France); Cordier, J.J.; Gargiulo, L. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Palmer, J.D. [EFDA CSU Garching (Germany); Martin, E. [ITER International Team, Garching (Germany); Tesini, A. [ITER International Team, ITER Naka Joint Work Site, Iberaki-ken (Japan)

    2004-07-01

    To perform an intervention a short time after plasma shutdown, the operation of the robot will have to be under ITER conditions which means: under high vacuum, pollution avoidance and a temperature ambience around 120 C. The feasibility studies have led to the design of a robot in the shape of a 8.2 meter long articulated arm made up of 5 modules with 11 articulated joints. A single module prototype has been manufactured to be tested. The prototype was set up in a specific vacuum vessel at Tore-Supra facility that can be baked up to 230 C under high-vacuum conditions. The first tests have shown that: -) the efficiency of the actuators at 120 C was the same than in air at room temperature, the speed was slightly lower, -) the monitoring of the temperature of the motor and of the power electronics components showed an increasing of only 40 C during 3 full pitch movements, and -) most of the greases was degassed during the 1 week long baking at 200 C except one which comes from an organic material in a component that has to be identified.

  4. ITER articulated inspection arm (AIA): R and D progress on vacuum and temperature technology for remote handling

    International Nuclear Information System (INIS)

    Perrot, Y.; Friconneau, J.P.; Cordier, J.J.; Gargiulo, L.; Martin, E.; Tesini, A.

    2004-01-01

    To perform an intervention a short time after plasma shutdown, the operation of the robot will have to be under ITER conditions which means: under high vacuum, pollution avoidance and a temperature ambience around 120 C. The feasibility studies have led to the design of a robot in the shape of a 8.2 meter long articulated arm made up of 5 modules with 11 articulated joints. A single module prototype has been manufactured to be tested. The prototype was set up in a specific vacuum vessel at Tore-Supra facility that can be baked up to 230 C under high-vacuum conditions. The first tests have shown that: -) the efficiency of the actuators at 120 C was the same than in air at room temperature, the speed was slightly lower, -) the monitoring of the temperature of the motor and of the power electronics components showed an increasing of only 40 C during 3 full pitch movements, and -) most of the greases was degassed during the 1 week long baking at 200 C except one which comes from an organic material in a component that has to be identified

  5. Manufacturing, assembly and tests of SPIDER Vacuum Vessel to develop and test a prototype of ITER neutral beam ion source

    Energy Technology Data Exchange (ETDEWEB)

    Zaccaria, Pierluigi, E-mail: pierluigi.zaccaria@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Masiello, Antonio [Fusion for Energy F4E, Barcelona (Spain); Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario [Ettore Zanon S.p.A., Schio (VI) (Italy)

    2015-10-15

    Highlights: • The SPIDER experiment aims to qualify and optimize the ion source for ITER injectors. • The large SPIDER Vacuum Vessel was built and it is under testing at the supplier. • The main working and assembly steps for production are presented in the paper. - Abstract: The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances.

  6. Manufacturing, assembly and tests of SPIDER Vacuum Vessel to develop and test a prototype of ITER neutral beam ion source

    International Nuclear Information System (INIS)

    Zaccaria, Pierluigi; Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco; Masiello, Antonio; Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario

    2015-01-01

    Highlights: • The SPIDER experiment aims to qualify and optimize the ion source for ITER injectors. • The large SPIDER Vacuum Vessel was built and it is under testing at the supplier. • The main working and assembly steps for production are presented in the paper. - Abstract: The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances.

  7. Concept development for the ITER equatorial port visible/infrared wide angle viewing system

    International Nuclear Information System (INIS)

    Reichle, R.; Beaumont, B.; Boilson, D.; Bouhamou, R.; Direz, M.-F.; Encheva, A.; Henderson, M.; Kazarian, F.; Lamalle, Ph.; Lisgo, S.; Mitteau, R.; Patel, K. M.; Pitcher, C. S.; Pitts, R. A.; Prakash, A.; Raffray, R.; Schunke, B.; Snipes, J.; Diaz, A. Suarez; Udintsev, V. S.

    2012-01-01

    The ITER equatorial port visible/infrared wide angle viewing system concept is developed from the measurement requirements. The proposed solution situates 4 viewing systems in the equatorial ports 3, 9, 12, and 17 with 4 views each (looking at the upper target, the inner divertor, and tangentially left and right). This gives sufficient coverage. The spatial resolution of the divertor system is 2 times higher than the other views. For compensation of vacuum-vessel movements, an optical hinge concept is proposed. Compactness and low neutron streaming is achieved by orienting port plug doglegs horizontally. Calibration methods, risks, and R and D topics are outlined.

  8. 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

  9. Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Lasnier, C. J., E-mail: lasnier@LLNL.gov; Allen, S. L.; Ellis, R. E.; Fenstermacher, M. E.; McLean, A. G.; Meyer, W. H.; Morris, K.; Seppala, L. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551-0808 (United States); Crabtree, K. [College of Optics, University of Arizona, Tucson, Arizona 85721 (United States); Van Zeeland, M. A. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)

    2014-11-15

    An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in diverted and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. Demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.

  10. Design of the ITER Tokamak Assembly Tools

    International Nuclear Information System (INIS)

    Park, Hyunki; Her, Namil; Kim, Byungchul; Im, Kihak; Jung, Kijung; Lee, Jaehyuk; Im, Kisuk

    2006-01-01

    ITER (International Thermonuclear Experimental Reactor) Procurement allocation among the seven Parties, EU, JA, CN, IN , KO, RF and US had been decided in Dec. 2005. ITER Tokamak assembly tools is one of the nine components allocated to Korea for the construction of the ITER. Assembly tools except measurement and common tools are supplied to assemble the ITER Tokamak and classified into 9 groups according to components to be assembled. Among the 9 groups of assembly tools, large-sized Sector Sub-assembly Tools and Sector Assembly Tools are used at the first stage of ITER Tokamak construction and need to be designed faster than seven other assembly tools. ITER IT (International Team) proposed Korea to accomplish ITA (ITER Transitional Arrangements) Task on detailed design, manufacturing feasibility and contract specification of specific, large sized tools such as Upending Tool, Lifting Tool, Sector Sub-assembly Tool and Sector Assembly Tool in Oct. 2004. Based on the concept design by ITER IT, Korea carried out ITA Task on detailed design of large-sized and specific Sector Sub-assembly and Sector Assembly Tools until Mar. 2006. The Sector Sub-assembly Tools mainly consist of the Upending, Lifting, Vacuum Vessel Support and Bracing, and Sector Sub-assembly Tool, among which the design of three tools are herein. The Sector Assembly Tools mainly consist of the Toroidal Field (TF) Gravity Support Assembly, Sector In-pit Assembly, TF Coil Assembly, Vacuum Vessel (VV) Welding and Vacuum Vessel Thermal Shield (TS) Assembly Tool, among which the design of Sector In-pit Assembly Tool is described herein

  11. Applicability assessment of plug weld to ITER vacuum vessel by crack propagation analysis

    International Nuclear Information System (INIS)

    Ohmori, Junji; Nakahira, Masataka; Takeda, Nobukazu; Shibanuma, Kiyoshi; Sago, Hiromi; Onozuka, Masanori

    2006-03-01

    In order to improve the fabricability of the vacuum vessel (VV) of International Thermonuclear Experimental Reactor (ITER), applicability of plug weld between VV outer shell and stiffening ribs/blanket support housings has been assessed using crack propagation analysis for the plug weld. The ITER VV is a double-wall structure of inner and outer shells with ribs and housings between the shells. For the fabrication of VV, ribs and housings are welded to outer shell after welding to inner shell. A lot of weld grooves should be adjusted for welding outer shell. The plug weld is that outer shells with slit at the weld region are set on ribs/housings then outer shells are welded to them by filling the slits with weld metal. The plug weld can allow larger tolerance of weld groove gap than ordinary butt weld. However, un-welded lengths parallel to outer sell surface remain in the plug weld region. It is necessary to evaluate the allowable un-welded length to apply the plug weld to ITER VV fabrication. For the assessment, the allowable un-welded lengths have been calculated by crack propagation analyses for load conditions, conservatively assuming the un-welded region is a crack. In the analyses, firstly allowable crack lengths are calculated from the stresses of the weld region. Then assuming initial crack length, crack propagation is calculated during operation period. Allowable initial crack lengths are determined on the condition that the propagated cracks should not exceed the allowable crack lengths. The analyses have been carried out for typical inboard straight region and inboard upper curved region with the maximum housing stress. The allowable initial cracks of ribs are estimated to be 8.8mm and 38mm for the rib and the housing, respectively, considering inspection error of 4.4mm. Plug weld between outer shell and ribs/housings could be applicable. (author)

  12. Fabrication of divertor cassette for ITER

    International Nuclear Information System (INIS)

    Sanguinetti, G.P.

    2008-01-01

    The Divertor is the component located on the bottom of the ITER vacuum vessel, whose main function is to adsorb the high thermal flux generated by the plasma whilst keeping the plasma impurity at a reasonable low level. The divertor consist of 54 units, each comprising outer components, facing the plasma and a component supporting the plasma facing components (PFC) and providing coolant distribution to them (divertor cassette). The divertor cassette is a box structure, butt welded and machined, made from plates and forgins of austenitic stainless steels. The cassette fabrication, which is in detail described, includes manufacturing of the attachments of the PFC to the cassette, the coolant distribution channels, and the cassette to vacuum vessel locking system. The divertor cassette is a pressure component (the cooling water runs at 40 bar) and therefore divertor cassette design, fabrication and service shall comply with the European PED and the applicable French law for the ITER. (orig.)

  13. CFD modelling of the wall friction velocity field in the ITER tokamak resulting from airflow during a loss of vacuum accident—Consequences for particle resuspension

    Energy Technology Data Exchange (ETDEWEB)

    Gélain, T., E-mail: thomas.gelain@irsn.fr; Rondeau, A.; Peillon, S.; Sabroux, J.C.; Gensdarmes, F.

    2015-11-15

    During a loss of vacuum accident (LOVA), dusts that will be present in the future tokamak ITER are likely to be re-suspended. Such aerosols formed may present a risk for explosion and airborne contamination. This article presents parameters that govern the forces affecting particles deposited on a wall and subject to airflow. It is shown the influence of three parameters in the dust mobilization mechanism, i.e.: the particles diameter, the fluid density and the friction velocity. From numerical simulations, it is determined the evolution of wall friction velocities in the vacuum vessel (VV) of ITER during a LOVA. The numerical calculations performed with ANSYS CFX code provide average friction velocities in the lower part of the tokamak between 12 m s{sup −1} at a pressure of 150 Pa, and 0.5 m s{sup −1} at a pressure of 10{sup 5} Pa.

  14. Design consideration on the synchrotron ultrahigh vacuum system

    International Nuclear Information System (INIS)

    Tsujikawa, H.; Chida, K.; Mizobuchi, A.; Miyahara, A.

    1982-01-01

    Ultrahigh vacuum production for the high-energy heavy-ion accelerator poses special problems concerning beam-gas molecule and beam-wall interactions. In this paper, summary of the TARN ultrahigh vacuum system and design criteria of the synchrotron ultrahigh vacuum system are presented. On-beam pressure of 4 x 10 -11 Torr is achieved in the TARN ultrahigh vacuum system, of which experiences through the construction and the operation are described and discussed. With emphasis on the application of newly developed technique in the fabrication of vacuum chamber and ultrahigh vacuum pump for the synchrotron ultrahigh vacuum system. (author)

  15. Electro-mechanical connection system for ITER in-vessel magnetic sensors

    Energy Technology Data Exchange (ETDEWEB)

    Rizzolo, Andrea; Brombin, Matteo; Gonzalez, Winder [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Marconato, Nicolò, E-mail: nicolo.marconato@igi.cnr.it [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Peruzzo, Simone [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Arshad, Shakeib [Fusion for Energy, C/Josep Pla, 2, 08019 Barcelona (Spain); Ma, Yunxing; Vayakis, George [ITER Organization, Route de Vinon-sur-Verdon, 13067 St Paul Lez Durance (France); Williams, Adrian [Oxford Technologies Ltd, 7 Nuffield Way, Abingdon, Oxon, OX14 1RL (United Kingdom)

    2016-11-01

    Highlights: • Latest status of the ITER “Generic In-Vessel Magnetic Platform” design activity. • Integration within the ITER In-Vessel configuration model. • Geometry optimization based on thermo-mechanical and magnetic field 3D calculation. • Assessment of the remote handling maintenance compatibility. - Abstract: This paper presents the preliminary design of the “In-Vessel Magnetic platform”, which is a subsystem of the magnetic diagnostics formed by all the components necessary for guaranteeing the thermo-mechanical interface of the actual magnetic sensors with the vacuum vessel (VV), their protection and the electrical connection to the in-vessel wiring for the transmission of the detected signal with a minimum level of noise. The design has been developed in order to comply with different functional requirements: the mechanical attachment to the VV; the electrical connection to the in-vessel wiring; efficient heat transfer to the VV; the compatibility with Remote Handling (RH) system for replacement; the integration of metrology features for post-installation control; the Electro Magnetic Interference (EMI) shielding from Electron Cyclotron Heating (ECH) stray radiation without compromising the sensor pass band (15 kHz). Significant effort has been dedicated to develop the CAD model, integrated within the ITER In-Vessel configuration model, taking care of the geometrical compliance with the Blanket modules (modified in order to accommodate the magnetic sensors in suitable grooves) and the RH compatibility. Thorough thermo-mechanical and electro-magnetic Finite Element Method (FEM) analyses have been performed to assess the reliability of the system in standard and off-normal operating conditions for the low frequency magnetic sensors.

  16. The JET ITER-like wall experiment: First results and lessons for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Horton, Lorne, E-mail: Lorne.Horton@jet.efda.org [EFDA-CSU Culham, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); European Commission, B-1049 Brussels (Belgium)

    2013-10-15

    Highlights: ► JET has recently completed the installation of an ITER-like wall. ► Important operational aspects have changed with the new wall. ► Initial experiments have confirmed the expected low fuel retention. ► Disruption dynamics have change dramatically. ► Development of wall-compatible, ITER-relevant regimes of operation has begun. -- Abstract: The JET programme is strongly focused on preparations for ITER construction and exploitation. To this end, a major programme of machine enhancements has recently been completed, including a new ITER-like wall, in which the plasma-facing armour in the main vacuum chamber is beryllium while that in the divertor is tungsten—the same combination of plasma-facing materials foreseen for ITER. The goal of the initial experimental campaigns is to fully characterise operation with the new wall, concentrating in particular on plasma-material interactions, and to make direct comparisons of plasma performance with the previous, carbon wall. This is being done in a progressive manner, with the input power and plasma performance being increased in combination with the commissioning of a comprehensive new real-time protection system. Progress achieved during the first set of experimental campaigns with the new wall, which took place from September 2011 to July 2012, is reported.

  17. Development and test of prototype components for ITER; Entwicklung und Test von Prototypkomponenten fuer ITER

    Energy Technology Data Exchange (ETDEWEB)

    Biel, Wolfgang; Behr, Wilfried; Castano-Bardawil, David; and others

    2015-08-15

    The scientific program of the project is divided into the following partial projects: (1.) ITER Diagnostic Port Plug for the charge-exchange spectroscopy (CXRS) with the subthemes: (a) Development of prototypes for critical mechanical components, (b) development of a roboter for the laser welding of vacuum seals and pipings at the Port Plug, (c) mirror studies, (d) CXRS prototype spectrometer, (2.) ITER tritium retention diagnostics (TR), (3.) ITER disruption mitigation ventile (DMV).

  18. Validation of the inspections with ultrasound of the welds of the reactor of ITER vacuum vessel; Validacion de las inspecciones con ultrasonidos de las soldaduras de la Vasija de Vacio del reactor del ITER

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, A.; Fernandez, F.; Perez, C.; Sillero, J. A.

    2013-07-01

    The ITER fusion reactor vacuum vessel has thousands of welding austenitic with shapes and different manufacturing processes. The RCC-MR code, which is that applied to the manufacture of the fusion reactor, requires a volumetric test all of them. This test should be mainly by x-rays and welds where it was not possible to use this method, ultrasonic.09-06.

  19. ITER Neutral Beam Injection System

    International Nuclear Information System (INIS)

    Ohara, Yoshihiro; Tanaka, Shigeru; Akiba, Masato

    1991-03-01

    A Japanese design proposal of the ITER Neutral Beam Injection System (NBS) which is consistent with the ITER common design requirements is described. The injection system is required to deliver a neutral deuterium beam of 75MW at 1.3MeV to the reactor plasma and utilized not only for plasma heating but also for current drive and current profile control. The injection system is composed of 9 modules, each of which is designed so as to inject a 1.3MeV, 10MW neutral beam. The most important point in the design is that the injection system is based on the utilization of a cesium-seeded volume negative ion source which can produce an intense negative ion beam with high current density at a low source operating pressure. The design value of the source is based on the experimental values achieved at JAERI. The utilization of the cesium-seeded volume source is essential to the design of an efficient and compact neutral beam injection system which satisfies the ITER common design requirements. The critical components to realize this design are the 1.3MeV, 17A electrostatic accelerator and the high voltage DC acceleration power supply, whose performances must be demonstrated prior to the construction of ITER NBI system. (author)

  20. 21 CFR 884.5070 - Vacuum abortion system.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Vacuum abortion system. 884.5070 Section 884.5070 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED... § 884.5070 Vacuum abortion system. (a) Identification. A vacuum abortion system is a device designed to...

  1. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, Masataka; Takeda, Nobukazu; Onozuka, Masanori [Japan Atomic Energy Agency (Japan); Kakudate, Satoshi [Mitsubishi Heavy Industries, Ltd. (Japan)

    2007-07-01

    The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. In the mock-up fabrication, it is planned to conduct the following items: 1. Feasibility of the Japanese proposed VV structure and fabrication methods and the applicability to the ITER are to be confirmed; 2. Assembly procedure and inspection procedure are to be confirmed; 3. Manufacturing tolerances are to be assessed; 4. Manufacturing schedule is to be assessed. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. General scope of the mock-up fabrication and the detailed dimensions are also shown. In the VV fabrication, several types of weld joint configuration will be used. This report shows the joint configurations proposed by Japan to be used for the inner shell connection, the rib-to-shell connection and outer shell connection, and the housing-to-shell connection, respectively. Non-destructive testing considered to be applied to each joint configuration is also presented. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of nondestructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. (orig.)

  2. Software design of the hybrid robot machine for ITER vacuum vessel assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Li, Ming, E-mail: Ming.Li@lut.fi [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Wu, Huapeng; Handroos, Heikki [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Yang, Guangyou [School of Mechanical Engineering, Hubei University of Technology, Wuhan (China)

    2013-10-15

    A specific software design is elaborated in this paper for the hybrid robot machine used for the ITER vacuum vessel (VV) assembly and maintenance. In order to provide the multi-machining-function as well as the complicated, flexible and customizable GUI designing satisfying the non-standardized VV assembly process in one hand, and in another hand guarantee the stringent machining precision in the real-time motion control of robot machine, a client–server-control software architecture is proposed, which separates the user interaction, data communication and robot control implementation into different software layers. Correspondingly, three particular application protocols upon the TCP/IP are designed to transmit the data, command and status between the client and the server so as to deal with the abundant data streaming in the software. In order not to be affected by the graphic user interface (GUI) modification process in the future experiment in VV assembly working field, the real-time control system is realized as a stand-alone module in the architecture to guarantee the controlling performance of the robot machine. After completing the software development, a milling operation is tested on the robot machine, and the result demonstrates that both the specific GUI operability and the real-time motion control performance could be guaranteed adequately in the software design.

  3. Software design of the hybrid robot machine for ITER vacuum vessel assembly and maintenance

    International Nuclear Information System (INIS)

    Li, Ming; Wu, Huapeng; Handroos, Heikki; Yang, Guangyou

    2013-01-01

    A specific software design is elaborated in this paper for the hybrid robot machine used for the ITER vacuum vessel (VV) assembly and maintenance. In order to provide the multi-machining-function as well as the complicated, flexible and customizable GUI designing satisfying the non-standardized VV assembly process in one hand, and in another hand guarantee the stringent machining precision in the real-time motion control of robot machine, a client–server-control software architecture is proposed, which separates the user interaction, data communication and robot control implementation into different software layers. Correspondingly, three particular application protocols upon the TCP/IP are designed to transmit the data, command and status between the client and the server so as to deal with the abundant data streaming in the software. In order not to be affected by the graphic user interface (GUI) modification process in the future experiment in VV assembly working field, the real-time control system is realized as a stand-alone module in the architecture to guarantee the controlling performance of the robot machine. After completing the software development, a milling operation is tested on the robot machine, and the result demonstrates that both the specific GUI operability and the real-time motion control performance could be guaranteed adequately in the software design

  4. The AGS Booster vacuum systems

    International Nuclear Information System (INIS)

    Hseuh, H.C.

    1989-01-01

    The AGS Booster is a synchrotron for the acceleration of both protons and heavy ions. The design pressure of low 10 -11 mbar is required to minimize beam loss of the partially stripped heavy ions. To remove contaminants and to reduce outgassing, the vacuum chambers and the components located in them will be chemically cleaned, vacuum fired, baked then treated with nitric oxide. The vacuum sector will be insitu baked to a minimum of 200 degree C and pumped by the combination of sputter ion pumps and titanium sublimation pumps. This paper describes the design and the processing of this ultra high vacuum system, and the performance of some half-cell vacuum chambers. 9 refs., 7 figs

  5. Concatenated coding system with iterated sequential inner decoding

    DEFF Research Database (Denmark)

    Jensen, Ole Riis; Paaske, Erik

    1995-01-01

    We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder......We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder...

  6. ITMETH, Iterative Routines for Linear System

    International Nuclear Information System (INIS)

    Greenbaum, A.

    1989-01-01

    1 - Description of program or function: ITMETH is a collection of iterative routines for solving large, sparse linear systems. 2 - Method of solution: ITMETH solves general linear systems of the form AX=B using a variety of methods: Jacobi iteration; Gauss-Seidel iteration; incomplete LU decomposition or matrix splitting with iterative refinement; diagonal scaling, matrix splitting, or incomplete LU decomposition with the conjugate gradient method for the problem AA'Y=B, X=A'Y; bi-conjugate gradient method with diagonal scaling, matrix splitting, or incomplete LU decomposition; and ortho-min method with diagonal scaling, matrix splitting, or incomplete LU decomposition. ITMETH also solves symmetric positive definite linear systems AX=B using the conjugate gradient method with diagonal scaling or matrix splitting, or the incomplete Cholesky conjugate gradient method

  7. Laser cleaning of ITER's diagnostic mirrors

    Science.gov (United States)

    Skinner, C. H.; Gentile, C. A.; Doerner, R.

    2012-10-01

    Practical methods to clean ITER's diagnostic mirrors and restore reflectivity will be critical to ITER's plasma operations. We report on laser cleaning of single crystal molybdenum mirrors coated with either carbon or beryllium films 150 - 420 nm thick. A 1.06 μm Nd laser system provided 220 ns pulses at 8 kHz with typical power densities of 1-2 J/cm^2. The laser beam was fiber optically coupled to a scanner suitable for tokamak applications. The efficacy of mirror cleaning was assessed with a new technique that combines microscopic imaging and reflectivity measurements [1]. The method is suitable for hazardous materials such as beryllium as the mirrors remain sealed in a vacuum chamber. Excellent restoration of reflectivity for the carbon coated Mo mirrors was observed after laser scanning under vacuum conditions. For the beryllium coated mirrors restoration of reflectivity has so far been incomplete and modeling indicates that a shorter duration laser pulse is needed. No damage of the molybdenum mirror substrates was observed.[4pt][1] C.H. Skinner et al., Rev. Sci. Instrum. at press.

  8. Particle contamination in vacuum systems

    International Nuclear Information System (INIS)

    Martignac, J.; Bonin, B.; Henriot, C.; Poupeau, J.P.; Koltchakian, I.; Kocic, D.; Herbeaux, Ch.; Marx, J.P.

    1996-01-01

    Many vacuum devices, like RF cavities, are sensitive to particle contamination. This fact has motivated a considerable effort of cleanliness from the SRF community. The present paper reports the first results of a general study trying to identify the most contaminating steps during assembly and vacuum operation of the cavity. The steps investigated here are gasket assembly, evacuation and venting of the vacuum system, and operation of sputter ion pumps. (author)

  9. Particle contamination in vacuum systems

    International Nuclear Information System (INIS)

    Martignac, J.; Bonin, B.; Henriot, C.; Poupeau, J.P.; Koltchakian, I.; Kocic, D.; Herbeaux, Ch.; Marx, J.P.

    1996-01-01

    Many vacuum devices, like RF cavities, are sensitive to particle contamination. This fact has motivated a considerable effort of cleanliness from the SRF community. The first results of a general study trying to identify the most contaminating steps during assembly and vacuum operation of the cavity is reported. The steps investigated here are gasket assembly, evacuation and venting of the vacuum system, and operation of sputter ion pumps. (author)

  10. Design foundation of vacuum system for electron beam machine

    International Nuclear Information System (INIS)

    Darsono; Suprapto; Djasiman

    1999-01-01

    Vacuum system is a main part of electron beam Machine because (EBM) the electron can not be produced without this vacuum. Vacuum system consists of vacuum pump, connecting pipe, valve, and vacuum gauge. The design vacuum system of EBM, basis knowledge and technology of vacuum is needed. The paper describes types of vacuum pump, calculation of pipe conductance and pumping time of vacuum system then there are used as consideration of criteria to choose vacuum pump for EBM. From the result of study, it is concluded that for EBM of 500 keV/10 mA which is going to use for wood coating and with consideration of economic and technic factor it is better to use diffusion pump. (author)

  11. 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.

  12. ITER ISS system alternative specification study

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-08-01

    Recent comments suggested that the fuel systems, in particular the ISS, could be simplified if the ITER specifications were relaxed from the data specified for ITER. This interim report addresses the first part of the analysis, which considers the impact of design specifications on fuel systems design

  13. ITER Dynamic Tritium Inventory Modeling Code

    International Nuclear Information System (INIS)

    Cristescu, Ioana-R.; Doerr, L.; Busigin, A.; Murdoch, D.

    2005-01-01

    A tool for tritium inventory evaluation within each sub-system of the Fuel Cycle of ITER is vital, with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems, however tritium accounting may be achieved by modeling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the systems. To get reliable results, an accurate dynamic modeling of the tritium content in each sub-system is necessary. In order to optimize the configuration and operation of the ITER fuel cycle, a dynamic fuel cycle model was developed progressively in the decade up to 2000-2001. As the design for some sub-systems from the fuel cycle (i.e. Vacuum pumping, Neutral Beam Injectors (NBI)) have substantially progressed meanwhile, a new code developed under a different platform to incorporate these modifications has been developed. The new code is taking over the models and algorithms for some subsystems, such as Isotope Separation System (ISS); where simplified models have been previously considered, more detailed have been introduced, as for the Water Detritiation System (WDS). To reflect all these changes, the new code developed inside EU participating team was nominated TRIMO (Tritium Inventory Modeling), to emphasize the use of the code on assessing the tritium inventory within ITER

  14. ELETTRA vacuum system

    International Nuclear Information System (INIS)

    Bernardini, M.; Daclon, F.; Giacuzzo, F.; Miertusova, J.; Pradal, F.; Kersevan, R.

    1993-01-01

    Elettra is a third-generation synchrotron light source which is being built especially for the use of high brilliance radiation from insertion devices and bending magnets. The UHV conditions in a storage ring lead to a longer beam lifetime - one of the most important criterion. The Elettra vacuum system presents some pecularities which cannot be found in any already existing machine. The final version of bending magnet vacuum chamber is presented. After chemical and thermal conditioning the specific outgassing rate of about 1.5e-12 Torr. liters sec -1 cm -2 was obtained. A microprocessor-controlled system has been developed to perform bake-out at the uniform temperature. The etched-foil type heaters are glued to the chamber and Microtherm insulation is used. UHV pumps based on standard triode sputter-ion pumps were modified with ST 707 NEG (Non Evaporable Getter) modules. A special installation enables the resistive activation of getters and significantly increases pumping speed for hydrogen and other residual gases (except methane and argon). All these technological innovations improve vacuum conditions in Elettra storage ring and consequently also the other parameters of the light source

  15. Experimental confirmation of the ITER cryopump high temperature regeneration scheme

    International Nuclear Information System (INIS)

    Day, C.; Haas, H.

    2007-01-01

    Forschungszentrum Karlsruhe (FZK) is developing the ITER high vacuum pumping systems for evacuation and maintenance of the required pressure levels in the torus (during burn and dwell, conditioning and leak detection), the neutral beam injectors and the cryostat vessel. All ITER high vacuum systems share the same concept of accumulative cryosorption pumping. The pumping surfaces, forced-cooled by 4.5 K supercritical helium, are coated with activated charcoal so as to be able to adsorb helium and hydrogens. All other gases are cryopumped by cryogenic phase transition from gaseous into the liquid/solid state. For the hydrogen processing pumps in the torus and the NBI, the maximum pumping time is given by the limitation of the maximum hydrogen inventory such that the resulting pressure in case of a loss of vacuum event and a corresponding oxy-hydrogen explosion is compatible to the design criteria of the vacuum vessel. To limit the gas accumulation, a staggered regeneration philosophy has been adopted, which involves three different temperature levels in order to achieve high regeneration efficiencies at best availability of the pumping system. The regular regeneration step is performed at a charcoal temperature of 90 K to release all hydrogen isotopomers (and helium), which are subsequently pumped out by the forevacuum pumping system. The second step at ambient temperature leads to the release of all air-like species. It has to be performed less frequently, probably over-night. Any water-like species with strong sorption bonding forces need still higher temperatures for effective desorption from the charcoal. These species comprise not only water itself but also high molecular tracers added to the water circuits in case of leak localisation and any pumped higher hydrocarbons from the plasma exhaust or. The latter in their tritiated forms may contribute significantly to the semi-permanent tritium inventory; a good knowledge of their regeneration characteristics is

  16. Disruptions, loads, and dynamic response of ITER

    International Nuclear Information System (INIS)

    Nelson, B.; Riemer, B.; Sayer, R.; Strickler, D.; Barabaschi, P.; Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

    1995-01-01

    Plasma disruptions and the resulting electromagnetic loads are critical to the design of the vacuum vessel and in-vessel components of the International Thermonuclear Experimental Reactor (ITER). This paper describes the status of plasma disruption simulations and related analysis, including the dynamic response of the vacuum vessel and in-vessel components, stresses and deflections in the vacuum vessel, and reaction loads in the support structures

  17. Diagnostic integration solutions in the ITER first wall

    International Nuclear Information System (INIS)

    Martínez, Gonzalo; Martin, Alex; Watts, Christopher; Veshchev, Evgeny; Reichle, Roger; Shigin, Pavel; Sabourin, Flavien; Gicquel, Stefan; Mitteau, Raphael; González, Jorge

    2015-01-01

    Highlights: • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets. • An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW. - Abstract: ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems.

  18. Diagnostic integration solutions in the ITER first wall

    Energy Technology Data Exchange (ETDEWEB)

    Martínez, Gonzalo, E-mail: gonzalo.martinez@iter.org [Technical University of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Martin, Alex; Watts, Christopher; Veshchev, Evgeny; Reichle, Roger [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Shigin, Pavel [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); National Research Nuclear University (MEPhI), Kashirskoe shosse, 115409 Moscow (Russian Federation); Sabourin, Flavien [ABMI-Groupe, Parc du Relais BatD 201 Route de SEDS, 13127 Vitrolles (France); Gicquel, Stefan; Mitteau, Raphael [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); González, Jorge [RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat (Spain)

    2015-10-15

    Highlights: • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets. • An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW. - Abstract: ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems.

  19. Procurement specification high vacuum test chamber and pumping system

    International Nuclear Information System (INIS)

    1976-01-01

    The specification establishes requirements for a high-vacuum test chamber, associated vacuum pumps, valves, controls, and instrumentation that shall be designed and fabricated for use as a test chamber for testing a closed loop Brayton Isotope Power System (BIPS) Ground Demonstration System (GDS). The vacuum system shall include all instrumentation required for pressure measurement and control of the vacuum pumping system. A general outline of the BIPS-GDS in the vacuum chamber and the preliminary piping and instrumentation interface to the vacuum chamber are shown

  20. Micro-particles in ITER: A comprehensive review

    International Nuclear Information System (INIS)

    Grisolia, C.; Rosanvallon, S.; Sharpe, Ph.; Winter, J.

    2009-01-01

    In a fusion reactor like ITER, in-vessel materials are subjected to interactions with the plasma. One of the main consequences of these plasma-material interactions is the creation of co-deposited layers. Due to internal stresses, part of these layers can crack leading to micro particle creation. The purpose of the following paper is to review the Tokamak operation processes which lead to erosion and layer creation. Then, the proportion of these layers that is converted into micro-particles will be evaluated in the case of Tore Supra experiments and extrapolated for ITER. It is major importance to measure the ITER mobilizable dusts present in the Vacuum Vessel and compare the measured quantity with the safety limits. When approaching these limits, removal systems must be used in order to control the in-vessel dust inventory. In the second part of the paper, diagnostics and removal system under development will be presented.

  1. CERN Vacuum-System Activities during the Long Shutdown 1: The LHC Beam Vacuum

    CERN Document Server

    Baglin, V; Chiggiato, P; Jimenez, JM; Lanza, G

    2014-01-01

    After the Long Shutdown 1 (LS1) and the consolidation of the magnet bus bars, the CERN Large Hadron Collider (LHC) will operate with nominal beam parameters. Larger beam energy, beam intensities and luminosity are expected. Despite the very good performance of the beam vacuum system during the 2010-12 physics run (Run 1), some particular areas require attention for repair, consolidation and upgrade. Among the main activities, a large campaign aiming at the repair of the RF bridges of some vacuum modules is conducted. Moreover, consolidation of the cryogenic beam vacuum systems with burst disk for safety reasons is implemented. In addition, NEG cartridges, NEG coated inserts and new instruments for the vacuum system upgrade are installed. Besides these activities, repair, consolidation and upgrades of other beam equipment such as collimators, kickers and beam instrumentations are carried out. In this paper, the motivation and the description for such activities, together with the expected beam vacuum performa...

  2. The ITER Thomson scattering core LIDAR diagnostic

    NARCIS (Netherlands)

    Naylor, G.A.; Scannell, R.; Beurskens, M.; Walsh, M.J.; Pastor, I.; Donné, A.J.H.; Snijders, B.; Biel, W.; Meszaros, B.; Giudicotti, L.; Pasqualotto, R.; Marot, L.

    2012-01-01

    The central electron temperature and density of the ITER plasma may be determined by Thomson scattering. A LIDAR topology is proposed in order to minimize the port access required of the ITER vacuum vessel. By using a LIDAR technique, a profile of the electron temperature and density can be

  3. The ITER Remote Maintenance Management System

    International Nuclear Information System (INIS)

    Tesini, Alessandro; Rolfe, A.C.

    2009-01-01

    A major challenge for the ITER project is to develop and implement a Remote Maintenance System, which can deliver high Tokamak availability within the constraints of the overall ITER programme objectives. Much of the maintenance of ITER will be performed using remote handling methods and some with combined manual and remote activities working together. The organization and management of the ITER remote handling facilities will be of a scale unlike any other remote handling application in the world. The ITER remote handling design and procurement activities will require co-ordination and management across many different sites throughout the world. It will be a major challenge for the ITER project to ensure a consistent quality and technical approach in all of the contributing parties. To address this issue the IO remote handling team are implementing the ITER Maintenance Management Plan (IMMP) comprising an overarching document defining the policies and methodologies (ITER Remote Maintenance Management System or IMMS) and an associated ITER remote handling code of practise (IRHCOP). The IMMS will be in document form available as a pdf file or similar. The IRHCOP will be implemented as a web based application and will provide access to the central resource of the entire code of practise from any location in the world. The IRHCOP data library will be centrally controlled in order that users can be assured of the data relevance and authenticity. This paper will describe the overall approach being taken to deal with this challenge and go on to detail the structure and content of both the IMMS and the IRHCOP.

  4. Weld distortion prediction of the ITER Vacuum Vessel using Finite Element simulations

    Energy Technology Data Exchange (ETDEWEB)

    Caixas, Joan, E-mail: joan.caixas@f4e.europa.eu [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Guirao, Julio [Numerical Analysis Technologies, S. L., Marqués de San Esteban 52, Entlo, 33209 Gijon (Spain); Bayon, Angel; Jones, Lawrence; Arbogast, Jean François [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Barbensi, Andrea [Ansaldo Nucleare, Corso F.M. Perrone, 25, I-16152 Genoa (Italy); Dans, Andres [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Facca, Aldo [Mangiarotti, Pannellia di Sedegliano, I-33039 Sedegliano (UD) (Italy); Fernandez, Elena; Fernández, José [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Iglesias, Silvia [Numerical Analysis Technologies, S. L., Marqués de San Esteban 52, Entlo, 33209 Gijon (Spain); Jimenez, Marc; Jucker, Philippe; Micó, Gonzalo [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Ordieres, Javier [Numerical Analysis Technologies, S. L., Marqués de San Esteban 52, Entlo, 33209 Gijon (Spain); Pacheco, Jose Miguel [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Paoletti, Roberto [Walter Tosto, Via Erasmo Piaggio, 72, I-66100 Chieti Scalo (Italy); Sanguinetti, Gian Paolo [Ansaldo Nucleare, Corso F.M. Perrone, 25, I-16152 Genoa (Italy); Stamos, Vassilis [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Tacconelli, Massimiliano [Walter Tosto, Via Erasmo Piaggio, 72, I-66100 Chieti Scalo (Italy)

    2013-10-15

    Highlights: ► Computational simulations of the weld processes can rapidly assess different sequences. ► Prediction of welding distortion to optimize the manufacturing sequence. ► Accurate shape prediction after each manufacture phase allows to generate modified procedures and pre-compensate distortions. ► The simulation methodology is improved using condensed computation techniques with ANSYS in order to reduce computation resources. ► For each welding process, the models are calibrated with the results of coupons and mock-ups. -- Abstract: The as-welded surfaces of the ITER Vacuum Vessel sectors need to be within a very tight tolerance, without a full-scale prototype. In order to predict welding distortion and optimize the manufacturing sequence, the industrial contract includes extensive computational simulations of the weld processes which can rapidly assess different sequences. The accurate shape prediction, after each manufacturing phase, enables actual distortions to be compared with the welding simulations to generate modified procedures and pre-compensate distortions. While previous mock-ups used heavy welded-on jigs to try to restrain the distortions, this method allows the use of lightweight jigs and yields important cost and rework savings. In order to enable the optimization of different alternative welding sequences, the simulation methodology is improved using condensed computation techniques with ANSYS in order to reduce computational resources. For each welding process, the models are calibrated with the results of coupons and mock-ups. The calibration is used to construct representative models of each segment and sector. This paper describes the application to the construction of the Vacuum Vessel sector of the enhanced simulation methodology with condensed Finite Element computation techniques and results of the calibration on several test pieces for different types of welds.

  5. Weld distortion prediction of the ITER Vacuum Vessel using Finite Element simulations

    International Nuclear Information System (INIS)

    Caixas, Joan; Guirao, Julio; Bayon, Angel; Jones, Lawrence; Arbogast, Jean François; Barbensi, Andrea; Dans, Andres; Facca, Aldo; Fernandez, Elena; Fernández, José; Iglesias, Silvia; Jimenez, Marc; Jucker, Philippe; Micó, Gonzalo; Ordieres, Javier; Pacheco, Jose Miguel; Paoletti, Roberto; Sanguinetti, Gian Paolo; Stamos, Vassilis; Tacconelli, Massimiliano

    2013-01-01

    Highlights: ► Computational simulations of the weld processes can rapidly assess different sequences. ► Prediction of welding distortion to optimize the manufacturing sequence. ► Accurate shape prediction after each manufacture phase allows to generate modified procedures and pre-compensate distortions. ► The simulation methodology is improved using condensed computation techniques with ANSYS in order to reduce computation resources. ► For each welding process, the models are calibrated with the results of coupons and mock-ups. -- Abstract: The as-welded surfaces of the ITER Vacuum Vessel sectors need to be within a very tight tolerance, without a full-scale prototype. In order to predict welding distortion and optimize the manufacturing sequence, the industrial contract includes extensive computational simulations of the weld processes which can rapidly assess different sequences. The accurate shape prediction, after each manufacturing phase, enables actual distortions to be compared with the welding simulations to generate modified procedures and pre-compensate distortions. While previous mock-ups used heavy welded-on jigs to try to restrain the distortions, this method allows the use of lightweight jigs and yields important cost and rework savings. In order to enable the optimization of different alternative welding sequences, the simulation methodology is improved using condensed computation techniques with ANSYS in order to reduce computational resources. For each welding process, the models are calibrated with the results of coupons and mock-ups. The calibration is used to construct representative models of each segment and sector. This paper describes the application to the construction of the Vacuum Vessel sector of the enhanced simulation methodology with condensed Finite Element computation techniques and results of the calibration on several test pieces for different types of welds

  6. Ultra high vacuum systems for accelerators

    International Nuclear Information System (INIS)

    Loefgren, P.

    2001-01-01

    Full text: In order to perform controlled, stable, and reproducible experiments, several research areas today require very low pressures. Maybe the most important example is the research that is performed in storage rings and accelerators where the lifetime and stability of particle beams depends critically on the vacuum conditions. Although the vacuum requirements ultimately depend on the kind of experiments that is performed, the studies of more and more rare and exotic species in storage rings and accelerators today pushes the demands on the vacuum conditions towards lower and lower pressures. The final pressure obtained in the vacuum system can often be the key factor for the outcome of an experiment. Pioneering work in vacuum technology has therefore often been performed at storage rings and accelerator facilities around the world. In order to reach pressures in the low UHV regime and lower (below 10 -11 mbar), several aspects have to be considered which implies choosing the proper materials, pumps and vacuum gauges. In the absence of gases inleaking from the outside, the rate of gas entering a vacuum system is determined by the release of molecules adsorbed on the surfaces and the outgassing from the bulk of the vacuum chamber walls. This means that the choice of material and, equally important, the pre treatment of the material, must be such that these rates are minimised. Today the most widely used material for vacuum applications are stainless steel. Besides its many mechanical advantages, it is resistant to corrosion and oxidation. If treated correctly the major gas source in a stainless steel chamber is hydrogen outgassing from the chamber walls. The hydrogen outgassing can be decreased by vacuum firing at 950 deg. C under vacuum. In addition to choosing the right materials the choice of vacuum pumps is important for the final pressure. Since no vacuum pump is capable of taking care of all kinds of gases found in the rest gas at pressures below 10 -11

  7. ITER lower port systems integration

    Energy Technology Data Exchange (ETDEWEB)

    Levesy, B., E-mail: bruno.levesy@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, D.; Boussier, B.; Bryan, S.; Cordier, J.J.; Dremel, M.; Dell' Orco, G.; Daly, E.; Doshi, B.; Jeannoutot, T.; Friconneau, J.P.; Gliss, C.; Le Barbier, R.; Lachevre, F.; Loughlin, M.; Martin, A.; Martins, J.P.; Maruyama, S.; Palmer, J.; Reichle, R. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2011-10-15

    The lower port systems are installed inside the vacuum vessel lower ports and in the adjacent port cells. The vacuum vessel ports and penetrations are allocated as follow: -4 ports dedicated to remote handling of the divertor cassettes, contain diagnostics racks and divertor cooling pipes. -5 ports connecting the main vessel to the torus cryopumps, contain divertor cooling pipes, pellet and gas injection pipes and vertical stabilization coil feeders. -3 penetrations connecting torus cryopumps are connected to the vacuum vessel by branch pipes. -Specific penetrations for divertor cooling lines, in-vessel viewing and glow discharge systems. The general layout of the port systems has been revised recently to improve the cryopump (8 t weight, 1.8 m diameter and 2.5 m long) maintenance scheme with remote handling tools and integrate the in-vessel vertical stabilization coil feeders. The port allocation, the pumping ports design, and interfaces in-between ports and cryostat and in-between cryopumps and cryostat have been up-dated. The integration inside the 18 port cells (11 m x 4 m each) has been reviewed to avoid clashes in between systems and to fix the openings in the port cell concrete walls. The new layout integrates safety and neutron-shielding requirements as well as remote handling and maintenance compatibility for the different systems. The paper presents the up-dated integration of the lower port systems inside the ports and the port cells. Interfaces of the port systems with the vacuum vessel, the cryostat and the port cells are described.

  8. 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.

  9. The remote handling systems for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Isabel, E-mail: mir@isr.ist.utl.pt [Institute for Systems and Robotics/Instituto Superior Tecnico, Lisboa (Portugal); Damiani, Carlo [Fusion for Energy, Barcelona (Spain); Tesini, Alessandro [ITER Organization, Cadarache (France); Kakudate, Satoshi [ITER Tokamak Device Group, Japan Atomic Energy Agency, Ibaraki (Japan); Siuko, Mikko [VTT Systems Engineering, Tampere (Finland); Neri, Carlo [Associazione EURATOM ENEA, Frascati (Italy)

    2011-10-15

    The ITER remote handling (RH) maintenance system is a key component in ITER operation both for scheduled maintenance and for unexpected situations. It is a complex collection and integration of numerous systems, each one at its turn being the integration of diverse technologies into a coherent, space constrained, nuclearised design. This paper presents an integrated view and recent results related to the Blanket RH System, the Divertor RH System, the Transfer Cask System (TCS), the In-Vessel Viewing System, the Neutral Beam Cell RH System, the Hot Cell RH and the Multi-Purpose Deployment System.

  10. Preliminary studies for the LHCb vertex detector vacuum system

    CERN Document Server

    Doets, M; Van Bakel, N; Van den Brand, J F J; van den Brand, Jo

    2000-01-01

    We lay down some general considerations which will serve as a starting point for design studies of a realistic LHCb vertex detector vacuum system. Based on these considerations, we propose a design strategy and identify issues to be further studied. In particular we try to outline some boundary conditions imposed by LHC and LHCb on the vacuum system. We discuss two possibilities for the LHCb vertex detector vacuum system. The preferred strategy uses a differentially pumped vacuum system with the silicon detectors separated from the beam line vacuum. Some estimations on static vacuum pressures and gas flows are presented.

  11. Vacuum systems for the ILC helical undulator

    CERN Document Server

    Malyshev, O B; Clarke, J A; Bailey, I R; Dainton, J B; Malysheva, L I; Barber, D P; Cooke, P; Baynham, E; Bradshaw, T; Brummitt, A; Carr, S; Ivanyushenkov, Y; Rochford, J; Moortgat-Pick, G A

    2007-01-01

    The International Linear Collider (ILC) positron source uses a helical undulator to generate polarized photons of ∼10MeV∼10MeV at the first harmonic. Unlike many undulators used in synchrotron radiation sources, the ILC helical undulator vacuum chamber will be bombarded by photons, generated by the undulator, with energies mostly below that of the first harmonic. Achieving the vacuum specification of ∼100nTorr∼100nTorr in a narrow chamber of 4–6mm4–6mm inner diameter, with a long length of 100–200m100–200m, makes the design of the vacuum system challenging. This article describes the vacuum specifications and calculations of the flux and energy of photons irradiating the undulator vacuum chamber and considers possible vacuum system design solutions for two cases: cryogenic and room temperature.

  12. Vacuum and fueling systems for the IGNITEX experiment

    International Nuclear Information System (INIS)

    Hallock, G.; Booth, W.D.; Carrera, R.

    1989-01-01

    The results of preliminary studies of the vacuum and fueling requirements for the proposed fusion ignition experiment IGNITEX are presented. An initial design for the vacuum pumping and plasma fueling system is given. The IGNITEX vacuum system must meet the demands of providing sufficient pumping speed to reach a base pressure of about 10 -8 Torr to provide a clean environment for plasma formation. In addition, the pumping speed should meet the requirements during the discharge cleaning cycle. The design of the vacuum pumping system including layout and location and structure of the vacuum ports required for pumping and diagnostic access is presented. Two different types of pumping systems - turbomolecular pumps and cryrogenic pumps have been considered. The advantages and disadvantages of each type of pumping system are analyzed

  13. Turbomolecular pump vacuum system for the Princeton Large Torus

    International Nuclear Information System (INIS)

    Dylla, H.F.

    1977-10-01

    A turbomolecular pump vacuum system has been designed and installed on the Princeton Large Torus (PLT). Four vertical shaft, oil-bearing, 1500 l/s turbomolecular pumps have been interfaced to the 6400 liter PLT Vacuum vessel to provide a net pumping speed of 3000 l/s for H 2 . The particular requirements and problems of tokamak vacuum systems are enumerated. A vacuum control system is described which protects the vacuum vessel from contamination, and protects the turbomolecular pumps from damage under a variety of possible failure modes. The performance of the vacuum system is presented in terms of pumping speed measurements and residual gas behavior

  14. Spent nuclear fuel project cold vacuum drying facility vacuum and purge system design description

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    1998-11-30

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Vacuum and Purge System (VPS) . The SDD was developed in conjunction with HNF-SD-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), The HNF-SD-SNF-DRD-002, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the VPS equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SDD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  15. Spent nuclear fuel project cold vacuum drying facility vacuum and purge system design description

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1998-01-01

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Vacuum and Purge System (VPS) . The SDD was developed in conjunction with HNF-SD-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), The HNF-SD-SNF-DRD-002, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the VPS equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SDD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  16. Optimization and Control for Sharing of the ITER Vacuum Vessel Support Force

    International Nuclear Information System (INIS)

    Rozov, V.

    2006-01-01

    The ITER Vacuum Vessel (VV) is a complex body supported in 9 points below lower ports by restraints in the radial, toroidal and vertical directions. The applied load produces a combination of reaction forces, which must be consistent with the design of the supported object. A reasonable sharing of the load among the supports is important for overall performance of the structure and helps to avoid excessive stress at the joints between the VV and lower ports. Optimization has been performed of the sharing of the total horizontal load applied to the ITER VV between radial and toroidal restraints. An effective method of finding simple parametric relationships between the design parameters of supports and the balance of the reaction forces has been developed. This allows purely analytical prediction of the sharing of the reaction forces for any desired stiffness of the applied restraints with no need for finite element structural analysis, and also allows control of the sharing by a proper selection of parameters of the supports. The method is based on the use of elementary mono-directional schemes - equivalent oscillators built for the main global modes, in static problems. The types of schemes and parameters of their members, related to the a-priori unknown stiffness of the VV structure under the supports, are found from consideration of the free vibration problem for the object using a 3D model of the VV with mass simulators - a series of simple eigenvalue analyses with variation of stiffness of the external restraints, that demands quite moderate computational resources. The equivalent schemes for the main modes not only enable simple one-line analytical calculation of the natural frequencies at any desired stiffness of the supports, but also indicate the contributions and balance of stiffness, to be considered in the static problem. The results of assessments of the reaction forces by direct static structural analyses for several cases are in agreement with values

  17. Design status of the ITER ECRH upper launcher mm-wave system

    Energy Technology Data Exchange (ETDEWEB)

    Landis, J.-D. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association Euratom-Confederation Suisse, 1015 Lausanne (Switzerland)], E-mail: jean-daniel.landis@epfl.ch; Chavan, R.; Bertizzolo, R.; Collazos, A.; Dolizy, F.; Felici, F.; Sanchez, F. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association Euratom-Confederation Suisse, 1015 Lausanne (Switzerland); Henderson, M. [ITER, Organization, Cadarache Centre, Saint Paul Lez Durance (France)

    2009-06-15

    The purpose of the ITER electron cyclotron resonance heating (ECRH) upper launcher (UL), or antennae will be to provide localised current drive by accurately directing mm-wave beams up to 2MW, out of the four allocated upper port plugs, at chosen rational magnetic flux surfaces in order to stabilise neoclassical tearing modes (NTMs). This paper will present an overview of the UL, with emphasis on the mm-wave components. The mm-wave layout includes corrugated waveguide sections and a quasi-optical path with both focusing mirrors and plane steering mirrors. One of the essential components of the UL is the Steering Mechanism Assembly (SMA), providing variable poloidal injection angles fulfilling high deposition accuracy requirements at the plasma location. The Actuator principle and rotor bearings are frictionless and backlash free, avoiding tribological difficulties such as stickslip and seizure. The underlying working principle is the use of mechanically compliant structures. Validation and proof testing of the steering principle is achieved with an uncooled first prototype demonstrator. A second prototype is currently being manufactured, comprising the functionalities needed for the ITER compatible system such as water cooling and high power mm-wave compatibility. In order to perform the fatigue tests of the actuator bellows, a test facility has been built, under ITER-like vacuum and temperature working conditions. Results of the cyclic fatigue tests are compared to the various manufacturer standards and codes, combining stress and strain controlled material fatigue properties.

  18. Comparison of vacuum rise time, vacuum limit accuracy, and occlusion break surge of 3 new phacoemulsification systems.

    Science.gov (United States)

    Han, Young Keun; Miller, Kevin M

    2009-08-01

    To compare vacuum rise time, vacuum limit accuracy, and occlusion break surge of 3 new phacoemulsification machines. Jules Stein Eye Institute and Department of Ophthalmology, David Geffen School of Medicine at UCLA, Los Angeles, California, USA. The vacuum rise time under normal and enhanced aspiration modes, vacuum limit accuracy, and occlusion break surge of the Infiniti Vision System, Stellaris Vision Enhancement System, and WhiteStar Signature Phacoemulsification System were tested. Vacuum rise time and limit accuracy were measured at limit settings of 400 mm Hg and 600 mm Hg. Surge area was recorded at vacuum limit settings of 200 mm Hg, 300 mm Hg, 400 mm Hg, and 500 mm Hg. The Infiniti had the fastest vacuum rise times under normal and enhanced aspiration modes. At 4 seconds, the vacuum limit accuracy was greatest with the Infiniti at the 400 mm Hg limit and the Signature at the 600 mm Hg limit. The Stellaris did not reach either vacuum target. The Infiniti performed better than the other 2 machines during testing of occlusion break surge at all vacuum limit settings above 200 mm Hg. Under controlled laboratory test conditions, the Infiniti had the fastest vacuum rise time, greatest vacuum limit accuracy at 400 mm Hg, and least occlusion break surge. These results can be explained by the lower compliance of the Infiniti system.

  19. HIGH PRODUCTIVITY VACUUM BLASTING SYSTEM

    International Nuclear Information System (INIS)

    McPhee, William S.

    1999-01-01

    The objective of this project is to improve the productivity and lower the expense of existing vacuum blasting technology. This technology is used to remove radioactive contamination, PCBs, and lead-based paint and provides worker protection by continuously recycling the material and dust for the decontamination tasks. The proposed work would increase the cleaning rate and provide safe and cost-effective decontamination of the DOE sites. This work focuses on redesigning and improving existing vacuum blasting technology including blast head nozzles, ergonomic handling of the blast head by reducing its weight; brush-ring design, vacuum level regulator, efficiency of the dust separator, and operational control sensors. The redesign is expected to enhance the productivity and economy of the vacuum blasting system by at least 50% over current vacuum blasting systems. There are three phases in the project. Phase I consists of developing and testing mathematical models. Phase II consists of pre-prototype design and fabrication and pre-prototype unit testing. Phase III consists of prototype design and field verification testing. In phase I, mathematical models are developed and analyzed for the nozzle, blast head, wind curtain, and dust separator, first as individual devices and then combined as an integrated model. This allows study of respective airflow and design parameters. The Contractor shall, based on the results of the mathematical modeling studies, design experimental models of the components and test these models. In addition, the Contractor shall develop sensors to detect the relationship of the blast head to the blast surfaces and controls to minimize the dependency on an operator's skill and judgment to obtain optimum positioning, as well as real-time characterization sensors to determine as the blast head is moving the depth to which coatings must be removed, thereby improving production and minimizing waste. In phase II, the Contractor shall design and

  20. Fabrication of full-size mock-up for 10° section of ITER vacuum vessel thermal shield

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Kwon [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo, E-mail: kwnam@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kang, Kyoung-O; Noh, Chang Hyun; Chung, Wooho [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Lim, Kisuk; Kang, Youngkil [SFA Engineering Corp., Asan-si, Chungcheongnam-do 336-873 (Korea, Republic of); Hamlyn-Harris, Craig; Her, Namil; Robby, Hicks [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    In this paper, a full-scale prototype fabrication for vacuum vessel thermal shield (VVTS) of ITER tokamak is described and test results are reported. All the manufacturing processes except for silver coating were performed in the fabrication of 10° section of VVTS. Pre-qualification test was conducted to compare the vertical and the horizontal welding positions. After shell welding, shell distortion was measured and adjusted. Shell thickness change was also measured after buffing process. Specially, VVTS ports need large bending and complex welding of shell and flange. Bending method for the complex and long cooling tube layout especially for the VVTS ports was developed in detail. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner and the scanning data was analyzed.

  1. ITER Remote Maintenance System (IRMS) lifecycle management

    Energy Technology Data Exchange (ETDEWEB)

    Tesini, Alessandro, E-mail: alessandro.tesini@iter.org [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Otto' , Bede [Oxford Technologies Ltd, 7, Nuffield Way, Abingdon, Oxon OX14 1RJ (United Kingdom); Blight, John [FAAST 31c Allee de la Granette, 13600 Ceyreste (France); Choi, Chang-Hwan; Friconneau, Jean-Pierre; Gotewal, Krishan Kumar; Hamilton, David [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Heckendorn, Frank [FD Technologies, PO Box 6686, Aiken, SC (United States); Martins, Jean-Pierre [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Marty, Thomas [Westinghouse, 122, avenue de Hambourg, 13008 Marseille (France); Nakahira, Masataka; Palmer, Jim; Subramanian, Rajendran [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2011-10-15

    The availability of the ITER machine to perform its scientific program is strongly dependent on the performance of the different Remote Handling (RH) systems constituting the ITER Remote Maintenance System (IRMS). The lifecycle of the IRMS will largely exceed 40 years from initial concept design and proof testing through to machine decommissioning. Such a long lifecycle requires that a rigorous approach is put in place to guarantee the technical capabilities of the highly innovative IRMS, its efficiency and its availability. For this purpose, an IRMS System Engineering and IRMS lifecycle management approach has been adopted by ITER. The approach aims at ensuring the IRMS full operability and availability at an acceptable cost of ownership over the full ITER machine assembly and operations period. The IRMS lifecycle management method described in this paper covers such subjects as specific requirements for IRMS design reviews, monitoring during manufacture, factory and site acceptance testing, integrated commissioning, decontamination, maintenance and re-qualification strategies, requirements for Integrated Logistical Support during operations. The updating and implementation of the IRMS lifecycle strategy and this procedure will be managed and monitored by the Remote Handling Integrated Product Team (RH-IPT). Although developed for the IRMS, the basic principles and procedures of lifecycle management could be applied to other ITER plant systems whose reliability and availability will be essential for the continued operation of the ITER machine.

  2. ITER Remote Maintenance System (IRMS) lifecycle management

    International Nuclear Information System (INIS)

    Tesini, Alessandro; Otto', Bede; Blight, John; Choi, Chang-Hwan; Friconneau, Jean-Pierre; Gotewal, Krishan Kumar; Hamilton, David; Heckendorn, Frank; Martins, Jean-Pierre; Marty, Thomas; Nakahira, Masataka; Palmer, Jim; Subramanian, Rajendran

    2011-01-01

    The availability of the ITER machine to perform its scientific program is strongly dependent on the performance of the different Remote Handling (RH) systems constituting the ITER Remote Maintenance System (IRMS). The lifecycle of the IRMS will largely exceed 40 years from initial concept design and proof testing through to machine decommissioning. Such a long lifecycle requires that a rigorous approach is put in place to guarantee the technical capabilities of the highly innovative IRMS, its efficiency and its availability. For this purpose, an IRMS System Engineering and IRMS lifecycle management approach has been adopted by ITER. The approach aims at ensuring the IRMS full operability and availability at an acceptable cost of ownership over the full ITER machine assembly and operations period. The IRMS lifecycle management method described in this paper covers such subjects as specific requirements for IRMS design reviews, monitoring during manufacture, factory and site acceptance testing, integrated commissioning, decontamination, maintenance and re-qualification strategies, requirements for Integrated Logistical Support during operations. The updating and implementation of the IRMS lifecycle strategy and this procedure will be managed and monitored by the Remote Handling Integrated Product Team (RH-IPT). Although developed for the IRMS, the basic principles and procedures of lifecycle management could be applied to other ITER plant systems whose reliability and availability will be essential for the continued operation of the ITER machine.

  3. Vacuum system of SST-1 Tokamak

    International Nuclear Information System (INIS)

    Khan, Ziauddin; Pathan, Firozkhan; George, Siju; Semwal, Pratibha; Dhanani, Kalpesh; Paravastu, Yuvakiran; Thankey, Prashant; Ramesh, Gattu; Himabindu, Manthena; Pradhan, Subrata

    2013-01-01

    Highlights: ► Air leaks developed during ongoing SST-1 cooldown campaign were detected online using RGA. ► The presence of N 2 and O 2 gases with the ratio of their partial pressures with ∼3.81:1 confirmed the air leaks. ► Baking of SST-1 was done efficiently by flowing hot N 2 gas in C-channels welded on inner surfaces without any problem. ► In-house fabricated demountable bull nose couplers were demonstrated for high temperature and pressure applications. ► Cryopumping effect was observed when liquid helium cooled superconducting magnets reached below 63 K. -- Abstract: Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN 2 cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10 −4 mbar and 1.0 × 10 −5 mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10 −6 mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10 −5 mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ∼3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10 −5 mbar is achieved inside the cryostat. Baking of the vacuum vessel up to 110 °C with ±10

  4. Physics design of the in-vessel collection optics for the ITER electron cyclotron emission diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Rowan, W. L., E-mail: w.l.rowan@austin.utexas.edu; Houshmandyar, S.; Phillips, P. E.; Austin, M. E. [Institute for Fusion Studies, The University of Texas at Austin, Austin, Texas 78712 (United States); Beno, J. H.; Ouroua, A. [Center for Electromechanics, The University of Texas at Austin, Austin, Texas 78712 (United States); Hubbard, A. E. [Plasma Science and Fusion Center, MIT, Cambridge, Massachusetts 02139 (United States); Khodak, A.; Taylor, G. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2016-11-15

    Measurement of the electron cyclotron emission (ECE) is one of the primary diagnostics for electron temperature in ITER. In-vessel, in-vacuum, and quasi-optical antennas capture sufficient ECE to achieve large signal to noise with microsecond temporal resolution and high spatial resolution while maintaining polarization fidelity. Two similar systems are required. One views the plasma radially. The other is an oblique view. Both views can be used to measure the electron temperature, while the oblique is also sensitive to non-thermal distortion in the bulk electron distribution. The in-vacuum optics for both systems are subject to degradation as they have a direct view of the ITER plasma and will not be accessible for cleaning or replacement for extended periods. Blackbody radiation sources are provided for in situ calibration.

  5. Integration of IC/EC systems in ITER

    International Nuclear Information System (INIS)

    Gassmann, T.; Beaumont, B.; Baruah, U.K.; Bonicelli, T.; Chiocchio, S.; Cox, D.; Darbos, C.; Decamps, H.; Denisov, G.; Henderson, M.; Kazarian, F.; Lamalle, P.U.; Mukherjee, A.; Rasmussen, D.; Saibene, G.; Sartori, R.; Sakamoto, K.; Tanga, A.

    2010-01-01

    The RF heating and current drive (H and CD) systems that are to be installed in ITER during the construction phase, are the electron cyclotron (EC) and ion cyclotron (IC) systems. They are complex assemblies of high voltage power supplies (HVPS), RF generators, transmission lines and antennas. Their design and integration are constrained by many interfaces, both internal, between the subsystems, and external, with the other ITER systems. In addition, some components must be compatible with a nuclear environment and are classified as Safety Important Component. This paper describes the processes implemented in ITER to ensure proper integration.

  6. Past, Present and Future of Linac 2 Vacuum System

    CERN Document Server

    Mahner, E

    2011-01-01

    This note aims to review the past, present, and future operation of CERN's Linac 2 vacuum system. The machine vacuum system layout with its major components is summarized. Operational problems arising in 2006 yielded to a leak test campaign of the whole machine, which detected a new, major leak on tank 3. Details about the mitigation of this leak are described as well as additional diagnostics installed during shutdown 2006/07. The pressure evolution of the most critical vacuum sector is analyzed. A statistics of vacuum system faults observed since 2000 is presented and compared with other systems. Finally, a perspective for the vacuum system operation until 2017/18 is outlined.

  7. Mechanical properties of electron beam welds of 316LN austenitic steels at low temperature for ITER gravity support system

    International Nuclear Information System (INIS)

    Lee, P.Y.; Huo, B.L.; Kuai, K.W.

    2007-01-01

    The gravity support system in ITER not only sustains magnet system, the vacuum vessel and in-vessel components, but also endures several large forces, such as electromagnetic force, thermal load and seismic loads. Based on the ITER design report, the maximum displacement of the gravity support system is estimated to be 32 mm in radial direction at the top flange of the flexible plates during the TF coil cool down from room temperature to 80 k. Welds are located in the peak stress region and subject to cyclic loads in the top flange is a potential problem. Therefore, the mechanical properties of the welds are extremely important for this system. 316LN austenitic stainless steel has been selected as the gravity support structure materials. However, there is still lack of the related mechanical data of the welding components of 316LN stainless steel at present. In this study, we are systematically investigated the mechanical properties of the welding components at low temperature. (authors)

  8. Development of an ITER relevant inspection robot

    Energy Technology Data Exchange (ETDEWEB)

    Gargiulo, Laurent [Association Euratom-CEA, Departement de Recherche sur la Fusion Controlee, CE Cadarache 13108 (France)], E-mail: laurent.gargiulo@cea.fr; Bayetti, Pascal; Bruno, Vincent; Cordier, Jean-Jacques [Association Euratom-CEA, Departement de Recherche sur la Fusion Controlee, CE Cadarache 13108 (France); Friconneau, Jean-Pierre [CEA-LIST Robotics and Interactive Systems Unit, CE Fontenay Aux Roses (France); Grisolia, Christian; Hatchressian, Jean-Claude; Houry, Michael [Association Euratom-CEA, Departement de Recherche sur la Fusion Controlee, CE Cadarache 13108 (France); Keller, Delphine; Perrot, Yann [CEA-LIST Robotics and Interactive Systems Unit, CE Fontenay Aux Roses (France)

    2008-12-15

    Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In particular, in-vessel inspection operations without loss of conditioning will be mandatory. In this context, an Articulated Inspection Arm (AIA) is currently developed by the CEA within the European work programme framework, which aims at demonstrating the feasibility of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier (up to 10 kg). It is composed of 5 segments with 8 degrees of freedom and a total range of 8 m. The first in situ tests will take place by the end of 2007 on the Tore Supra Tokamak at Cadarache (France). They will validate concepts for operations under ITER relevant vacuum and temperature conditions. After qualification, the arm will constitute a promising tool for various applications. Several processes are already considered for ITER maintenance and will be demonstrated on the AIA robot carrier: - The first embedded process is the viewing system. It is already manufactured and will allow close visual inspection of the complex Plasma Facing Components (PFC) (limiters, neutralisers, RF antenna, diagnostic windows, etc.). - In situ localisation of water leakage based on a helium sniffing system is also being studied to improve and facilitate maintenance operations. - Finally a laser ablation system for PFC detritiation, developed in CEA laboratories, is being fitted to be implemented on the robot for future operation in Tore Supra. This paper deals with the integration of the robot into Tore Supra and the progress in the development of the processes listed above. It also describes the current test campaign aiming to qualify the robot performance and reliability under vacuum and temperature conditions.

  9. Development of an ITER relevant inspection robot

    International Nuclear Information System (INIS)

    Gargiulo, Laurent; Bayetti, Pascal; Bruno, Vincent; Cordier, Jean-Jacques; Friconneau, Jean-Pierre; Grisolia, Christian; Hatchressian, Jean-Claude; Houry, Michael; Keller, Delphine; Perrot, Yann

    2008-01-01

    Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In particular, in-vessel inspection operations without loss of conditioning will be mandatory. In this context, an Articulated Inspection Arm (AIA) is currently developed by the CEA within the European work programme framework, which aims at demonstrating the feasibility of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier (up to 10 kg). It is composed of 5 segments with 8 degrees of freedom and a total range of 8 m. The first in situ tests will take place by the end of 2007 on the Tore Supra Tokamak at Cadarache (France). They will validate concepts for operations under ITER relevant vacuum and temperature conditions. After qualification, the arm will constitute a promising tool for various applications. Several processes are already considered for ITER maintenance and will be demonstrated on the AIA robot carrier: - The first embedded process is the viewing system. It is already manufactured and will allow close visual inspection of the complex Plasma Facing Components (PFC) (limiters, neutralisers, RF antenna, diagnostic windows, etc.). - In situ localisation of water leakage based on a helium sniffing system is also being studied to improve and facilitate maintenance operations. - Finally a laser ablation system for PFC detritiation, developed in CEA laboratories, is being fitted to be implemented on the robot for future operation in Tore Supra. This paper deals with the integration of the robot into Tore Supra and the progress in the development of the processes listed above. It also describes the current test campaign aiming to qualify the robot performance and reliability under vacuum and temperature conditions

  10. Vacuum system operating experience review for fusion applications

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1994-03-01

    This report presents a review of vacuum system operating experiences from particle accelerator, fusion experiment, space simulation chamber, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of vacuum system component failure rates and accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with vacuum systems are discussed, including personnel safety, foreign material intrusion, and factors relevant to vacuum systems being the primary confinement boundary for tritium and activated dusts. This information should be useful to fusion system designers and safety analysts, such as the team working on the Engineering Design Activities for the International Thermonuclear Experimental Reactor

  11. Characteristics of the ISABELLE vacuum system

    International Nuclear Information System (INIS)

    Aggus, J.R.; Edwards, D. Jr.; Halama, H.J.; Herrera, J.C.

    1977-01-01

    A discussion is given of the complete vacuum system of ISABELLE, emphasizing those design characteristics dictated by high vacuum, the avoidance of beam current loss, and the reduction of background. The experimental and theoretical justifications for the design are presented

  12. Status of Preliminary Design on the Assembly Tools for ITER Tokamak Machine

    International Nuclear Information System (INIS)

    Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin; Moon, Jae Hwan; Kim, Byung Seok; Lee, Jae Hyuk; Shaw, Robert

    2012-01-01

    The ITER Tokamak device is principally composed of nine 40 .deg. sectors. Each 40 .deg. sector is made up of one 40 .deg. vacuum vessel (VV), two 20 .deg. toroidal filed coils (TFC) and associated vacuum vessel thermal shield (VVTS) segments which consist of one inboard and two outboard vacuum vessel thermal shields. Based on the design description document and final report prepared by the ITER organization (IO) and conceptual design, Korea has carried out the preliminary design of these assembly tools. The assembly strategy and relevant tools for the 40 .deg. sector sub-assembly and sector assembly at in-pit should be developed to satisfy the basic assembly requirements of the ITER Tokamak machine. Assembly strategy, preliminary design of the sector sub-assembly and assembly tools are described in this paper

  13. Analytical and numerical tools for vacuum systems

    CERN Document Server

    Kersevan, R

    2007-01-01

    Modern particle accelerators have reached a level of sophistication which require a thorough analysis of all their sub-systems. Among the latter, the vacuum system is often a major contributor to the operating performance of a particle accelerator. The vacuum engineer has nowadays a large choice of computational schemes and tools for the correct analysis, design, and engineering of the vacuum system. This paper is a review of the different type of algorithms and methodologies which have been developed and employed in the field since the birth of vacuum technology. The different level of detail between simple back-of-the-envelope calculations and more complex numerical analysis is discussed by means of comparisons. The domain of applicability of each method is discussed, together with its pros and cons.

  14. Neutron cameras for ITER

    International Nuclear Information System (INIS)

    Johnson, L.C.; Barnes, C.W.; Batistoni, P.

    1998-01-01

    Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This paper analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, the authors examine proposed compact cameras, in which neutron fluxes are inferred from 16 N decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with 16 N-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins

  15. Vacuum system of SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Ziauddin, E-mail: ziauddin@ipr.res.in [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382 428 (India); Pathan, Firozkhan; George, Siju; Semwal, Pratibha; Dhanani, Kalpesh; Paravastu, Yuvakiran; Thankey, Prashant; Ramesh, Gattu; Himabindu, Manthena; Pradhan, Subrata [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382 428 (India)

    2013-10-15

    Highlights: ► Air leaks developed during ongoing SST-1 cooldown campaign were detected online using RGA. ► The presence of N{sub 2} and O{sub 2} gases with the ratio of their partial pressures with ∼3.81:1 confirmed the air leaks. ► Baking of SST-1 was done efficiently by flowing hot N{sub 2} gas in C-channels welded on inner surfaces without any problem. ► In-house fabricated demountable bull nose couplers were demonstrated for high temperature and pressure applications. ► Cryopumping effect was observed when liquid helium cooled superconducting magnets reached below 63 K. -- Abstract: Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN{sub 2} cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10{sup −4} mbar and 1.0 × 10{sup −5} mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10{sup −6} mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10{sup −5} mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ∼3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10{sup −5} mbar is achieved inside the cryostat. Baking of the

  16. 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.)

  17. Development and test of prototype components for ITER

    International Nuclear Information System (INIS)

    Biel, Wolfgang; Behr, Wilfried; Castano-Bardawil, David

    2015-08-01

    The scientific program of the project is divided into the following partial projects: (1.) ITER Diagnostic Port Plug for the charge-exchange spectroscopy (CXRS) with the subthemes: (a) Development of prototypes for critical mechanical components, (b) development of a roboter for the laser welding of vacuum seals and pipings at the Port Plug, (c) mirror studies, (d) CXRS prototype spectrometer, (2.) ITER tritium retention diagnostics (TR), (3.) ITER disruption mitigation ventile (DMV).

  18. Troubleshooting vacuum systems steam turbine surface condensers and refinery vacuum towers

    CERN Document Server

    Lieberman, Norman P

    2012-01-01

    Vacuum systems are in wide spread use in the petrochemical plants, petroleum refineries and power generation plants. The existing texts on this subject are theoretical in nature and only deal with how the equipment functions when in good mechanical conditions, from the viewpoint of the equipment vendor.  In this much-anticipated volume, one of the most well-respected and prolific process engineers in the world takes on troubleshooting vacuum systems, and especially steam ejectors, an extremely complex and difficult subject that greatly effects the profitability of the majority of the world'

  19. 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.)

  20. 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

  1. Challenging issues in the design and manufacturing of the European sectors of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Dans, Andres; Jucker, P.; Bayon, A.; Arbogast, J.-F.; Caixas, J.; Fernández, J.; Micó, G.; Pacheco, J.; Trentea, A.; Stamos, V.

    2014-01-01

    Highlights: • ITER Vacuum Vessel was described with its features and particularities. • Engineering and CAD design of Sector 5 is finish; the work of sectors 3 and 4 is ongoing. • Fabrication Mock Ups almost finished with an important know-how acquired. • Procurement of raw material (plates and forgings) started. • Qualification of welding, NDT and forming close to be finished. - Abstract: Fusion for Energy (F4E), the European Domestic Agency for the ITER project, has to supply seven sectors as part of the European contribution to the project. F4E signed the Procurement Agreement with ITER Organization (IO) in 2009. After a call for tender in 2010, the contract for the manufacturing of seven sectors was placed in October 2010 to a consortium of three Italian companies, Ansaldo, Mangiarotti and Walter Tosto (AMW). The first sector in the manufacturing route is Sector 5 (later will come 4, 3, 2, 9, 8, 7). This paper will cover: the status of the engineering activities, design, procurement and preparation to begin the manufacturing in 2013. Also will be presented the statutory and regulatory requirements of the French Nuclear Safety regulator and the status of the relevant R and D mock-ups to demonstrate manufacturing feasibility control of distortions (using predictions with analysis and algorithms to change in real time the manufacturing route in order to correct such distortions, inspectability and metrology). Another important aspect at this stage of the manufacturing is qualification of activities like welding, Non-destructive Examination and Hot Forming. This paper describes the status of the activities currently in process in order to meet with the challenging design, schedule and high quality requirements of the project

  2. Challenging issues in the design and manufacturing of the European sectors of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Dans, Andres, E-mail: andresdans@gmail.com; Jucker, P.; Bayon, A.; Arbogast, J.-F.; Caixas, J.; Fernández, J.; Micó, G.; Pacheco, J.; Trentea, A.; Stamos, V.

    2014-10-15

    Highlights: • ITER Vacuum Vessel was described with its features and particularities. • Engineering and CAD design of Sector 5 is finish; the work of sectors 3 and 4 is ongoing. • Fabrication Mock Ups almost finished with an important know-how acquired. • Procurement of raw material (plates and forgings) started. • Qualification of welding, NDT and forming close to be finished. - Abstract: Fusion for Energy (F4E), the European Domestic Agency for the ITER project, has to supply seven sectors as part of the European contribution to the project. F4E signed the Procurement Agreement with ITER Organization (IO) in 2009. After a call for tender in 2010, the contract for the manufacturing of seven sectors was placed in October 2010 to a consortium of three Italian companies, Ansaldo, Mangiarotti and Walter Tosto (AMW). The first sector in the manufacturing route is Sector 5 (later will come 4, 3, 2, 9, 8, 7). This paper will cover: the status of the engineering activities, design, procurement and preparation to begin the manufacturing in 2013. Also will be presented the statutory and regulatory requirements of the French Nuclear Safety regulator and the status of the relevant R and D mock-ups to demonstrate manufacturing feasibility control of distortions (using predictions with analysis and algorithms to change in real time the manufacturing route in order to correct such distortions, inspectability and metrology). Another important aspect at this stage of the manufacturing is qualification of activities like welding, Non-destructive Examination and Hot Forming. This paper describes the status of the activities currently in process in order to meet with the challenging design, schedule and high quality requirements of the project.

  3. 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

  4. Wireless Integrated Microelectronic Vacuum Sensor System

    Science.gov (United States)

    Krug, Eric; Philpot, Brian; Trott, Aaron; Lawrence, Shaun

    2013-01-01

    NASA Stennis Space Center's (SSC's) large rocket engine test facility requires the use of liquid propellants, including the use of cryogenic fluids like liquid hydrogen as fuel, and liquid oxygen as an oxidizer (gases which have been liquefied at very low temperatures). These fluids require special handling, storage, and transfer technology. The biggest problem associated with transferring cryogenic liquids is product loss due to heat transfer. Vacuum jacketed piping is specifically designed to maintain high thermal efficiency so that cryogenic liquids can be transferred with minimal heat transfer. A vacuum jacketed pipe is essentially two pipes in one. There is an inner carrier pipe, in which the cryogenic liquid is actually transferred, and an outer jacket pipe that supports and seals the vacuum insulation, forming the "vacuum jacket." The integrity of the vacuum jacketed transmission lines that transfer the cryogenic fluid from delivery barges to the test stand must be maintained prior to and during engine testing. To monitor the vacuum in these vacuum jacketed transmission lines, vacuum gauge readings are used. At SSC, vacuum gauge measurements are done on a manual rotation basis with two technicians, each using a handheld instrument. Manual collection of vacuum data is labor intensive and uses valuable personnel time. Additionally, there are times when personnel cannot collect the data in a timely fashion (i.e., when a leak is detected, measurements must be taken more often). Additionally, distribution of this data to all interested parties can be cumbersome. To simplify the vacuum-gauge data collection process, automate the data collection, and decrease the labor costs associated with acquiring these measurements, an automated system that monitors the existing gauges was developed by Invocon, Inc. For this project, Invocon developed a Wireless Integrated Microelectronic Vacuum Sensor System (WIMVSS) that provides the ability to gather vacuum

  5. Progress in the conceptual design of the ITER cask and plug remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Locke, Darren, E-mail: darren.locke@f4e.europa.eu [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); González Gutiérrez, Carmen; Damiani, Carlo [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Friconneau, Jean-Pierre; Martins, Jean-Pierre [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    Highlights: • The CPRHS is a complex system with a significant number of complicated interfaces. • Significant effort is being made to ensure that the system requirements are clearly defined. • This solution relates to planned operations and also anticipation of rescue operations. • With the CPRHS performing a safety function process control is being put in place. • All these factors will have a significant impact on the success of the CPRHS. - Abstract: One function of the ITER remote maintenance system is the transportation of in-vessel components and remote handling systems to and from the vacuum vessel and docking stations in the Hot Cell via dedicated galleries and lift. The cask and plug remote handling system (CPRHS) has been adopted as the solution to provide this nuclear confinement and transportation. This paper discusses the development of the conceptual design to-date and presents the processes being implemented to effectively control the subsequent CPRHS development. The CPRHS is a complex suite of systems with a significant number of interfaces with other ITER systems. Significant effort is being made to ensure that the system requirements are comprehensively defined and carefully managed and a feasible solution is developed – including planned and rescue operations. With the CPRHS performing a critical confinement function appropriate processes are being put in place to control the system development of the CPRHS. The expectation is that the combination of these factors will have a significant impact on the successful implementation of the CPRHS.

  6. Progress in the conceptual design of the ITER cask and plug remote handling system

    International Nuclear Information System (INIS)

    Locke, Darren; González Gutiérrez, Carmen; Damiani, Carlo; Friconneau, Jean-Pierre; Martins, Jean-Pierre

    2014-01-01

    Highlights: • The CPRHS is a complex system with a significant number of complicated interfaces. • Significant effort is being made to ensure that the system requirements are clearly defined. • This solution relates to planned operations and also anticipation of rescue operations. • With the CPRHS performing a safety function process control is being put in place. • All these factors will have a significant impact on the success of the CPRHS. - Abstract: One function of the ITER remote maintenance system is the transportation of in-vessel components and remote handling systems to and from the vacuum vessel and docking stations in the Hot Cell via dedicated galleries and lift. The cask and plug remote handling system (CPRHS) has been adopted as the solution to provide this nuclear confinement and transportation. This paper discusses the development of the conceptual design to-date and presents the processes being implemented to effectively control the subsequent CPRHS development. The CPRHS is a complex suite of systems with a significant number of interfaces with other ITER systems. Significant effort is being made to ensure that the system requirements are comprehensively defined and carefully managed and a feasible solution is developed – including planned and rescue operations. With the CPRHS performing a critical confinement function appropriate processes are being put in place to control the system development of the CPRHS. The expectation is that the combination of these factors will have a significant impact on the successful implementation of the CPRHS

  7. Large eddy simulation of Loss of Vacuum Accident in STARDUST facility

    International Nuclear Information System (INIS)

    Benedetti, Miriam; Gaudio, Pasquale; Lupelli, Ivan; Malizia, Andrea; Porfiri, Maria Teresa; Richetta, Maria

    2013-01-01

    Highlights: ► Fusion safety, plasma material interaction. ► Numerical and experimental data comparison to analyze the consequences of Loss of Vacuum Accident that can provoke dust mobilization inside the Vacuum Vessel of the Nuclear Fusion Reactor ITER-like. -- Abstract: The development of computational fluid dynamic (CFD) models of air ingress into the vacuum vessel (VV) represents an important issue concerning the safety analysis of nuclear fusion devices, in particular in the field of dust mobilization. The present work deals with the large eddy simulations (LES) of fluid dynamic fields during a vessel filling at near vacuum conditions to support the safety study of Loss of Vacuum Accidents (LOVA) events triggered by air income. The model's results are compared to the experimental data provided by STARDUST facility at different pressurization rates (100 Pa/s, 300 Pa/s and 500 Pa/s). Simulation's results compare favorably with experimental data, demonstrating the possibility of implementing LES in large vacuum systems as tokamaks

  8. Re-circulating linac vacuum system

    International Nuclear Information System (INIS)

    Wells, Russell P.; Corlett, John N.; Zholents, Alexander A.

    2003-01-01

    The vacuum system for a proposed 2.5 GeV, 10ΜA recirculating linac synchrotron light source [1] is readily achievable with conventional vacuum hardware and established fabrication processes. Some of the difficult technical challenges associated with synchrotron light source storage rings are sidestepped by the relatively low beam current and short beam lifetime requirements of a re-circulating linac. This minimal lifetime requirement leads directly to relatively high limits on the background gas pressure through much of the facility. The 10ΜA average beam current produces very little synchrotron radiation induced gas desorption and thus the need for an ante-chamber in the vacuum chamber is eliminated. In the arc bend magnets, and the insertion devices, the vacuum chamber dimensions can be selected to balance the coherent synchrotron radiation and resistive wall wakefield effects, while maintaining the modest limits on the gas pressure and minimal outgassing

  9. Development of an ITER relevant inspection robot

    Energy Technology Data Exchange (ETDEWEB)

    Gargiulo, L.; Bayetti, P.; Cordier, J.J.; Grisolia, C.; Hatchressian, J.C. [Association Euratom-CEA, Cadarache (France). Dept. de Recherche sur la Fusion Controlee; Friconneau, J.P.; Keller, D.; Perrot, Y. [CEA-LIST Robotics and Interactive Systems Unit, Fontenay aux Roses (France)

    2007-07-01

    Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In particular, in vessel inspection operations without loss of conditioning could be very useful. Within this framework, the aim of the project called AIA (Articulated Inspection Arm) is to demonstrate the feasibility of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier (up to 10 kg). It is composed of 5 segments with 11 degrees of freedom and a total range of 8 m. The project is currently developed by the CEA within the European workprogramme. Its first in situ tests are planned this summer on the Tore Supra tokamak at Cadarache (France). They will validate chosen concepts for operations under ITER relevant vacuum and temperature conditions. After qualification, the arm will constitute a promising tool for generic application. Several processes are already considered for ITER maintenance and will be demonstrated on the AIA robot carrier: - The first embedded process is the viewing system. It is currently being manufactured and will allow for close visual inspection of the complex Plasma Facing Components (limiters, neutralisers, RF antennae, diagnostic windows, etc.). - In situ localisation of leakage based on helium sniffer is also studied to improve maintenance operations. - Finally the laser ablation system for PFC detritiation, also developed in CEA laboratories, is being fitted to be implanted into the robot and put into operation in Tore Supra. This paper deals with the integration of the robot in the Tore Supra tokamak and the advances in the development of the listed processes. It also introduces the current test campaign aiming to qualify the robot performance and reliability under vacuum and temperature conditions. (orig.)

  10. Development of an ITER relevant inspection robot

    International Nuclear Information System (INIS)

    Gargiulo, L.; Bayetti, P.; Cordier, J.J.; Grisolia, C.; Hatchressian, J.C.

    2007-01-01

    Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In particular, in vessel inspection operations without loss of conditioning could be very useful. Within this framework, the aim of the project called AIA (Articulated Inspection Arm) is to demonstrate the feasibility of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier (up to 10 kg). It is composed of 5 segments with 11 degrees of freedom and a total range of 8 m. The project is currently developed by the CEA within the European workprogramme. Its first in situ tests are planned this summer on the Tore Supra tokamak at Cadarache (France). They will validate chosen concepts for operations under ITER relevant vacuum and temperature conditions. After qualification, the arm will constitute a promising tool for generic application. Several processes are already considered for ITER maintenance and will be demonstrated on the AIA robot carrier: - The first embedded process is the viewing system. It is currently being manufactured and will allow for close visual inspection of the complex Plasma Facing Components (limiters, neutralisers, RF antennae, diagnostic windows, etc.). - In situ localisation of leakage based on helium sniffer is also studied to improve maintenance operations. - Finally the laser ablation system for PFC detritiation, also developed in CEA laboratories, is being fitted to be implanted into the robot and put into operation in Tore Supra. This paper deals with the integration of the robot in the Tore Supra tokamak and the advances in the development of the listed processes. It also introduces the current test campaign aiming to qualify the robot performance and reliability under vacuum and temperature conditions. (orig.)

  11. Progress of IRSN R&D on ITER Safety Assessment

    Science.gov (United States)

    Van Dorsselaere, J. P.; Perrault, D.; Barrachin, M.; Bentaib, A.; Gensdarmes, F.; Haeck, W.; Pouvreau, S.; Salat, E.; Seropian, C.; Vendel, J.

    2012-08-01

    The French "Institut de Radioprotection et de Sûreté Nucléaire" (IRSN), in support to the French "Autorité de Sûreté Nucléaire", is analysing the safety of ITER fusion installation on the basis of the ITER operator's safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.

  12. Current status of ITER I&C system as integration begins

    Energy Technology Data Exchange (ETDEWEB)

    Davis, William, E-mail: william.davis@iter.org [ITER Organisation, Route de Vinon-sur Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Wallander, Anders [ITER Organisation, Route de Vinon-sur Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Yonekawa, Izuru [Nippon Advanced Technology Ltd., 3129-45 Hibara Muramatsu, Tokai, Naka-gun, Ibaraki 319-1112 (Japan)

    2016-11-15

    Highlights: • The ITER I&C system is organisationally complicated and technically challenging. • Standard technologies for the ITER I&C systems have been selected. • Supply of non-standard technologies will cause serious issues. • Differing levels of design maturity of plant I&C systems is a serious challenge. • Systems are in the final stages of design and are being delivered to site. - Abstract: The ITER I&C system is organisationally complicated and technically challenging, and integrating its many sub-systems into a single coherent system is critical for the ITER project to meet its objectives. This paper explains the integration risks being faced now and anticipated in the near future. Standardisation initiatives by the ITER central team to mitigate these risks are described. The paper also presents the architecture of the ITER I&C system, the current status of design and manufacture key developments made in recent years, and the current and future activities of the central I&C teams. Finally, a short description is given of the plant I&C systems that will be delivered to ITER in the near future.

  13. Vacuum system transient simulator and its application to TFTR

    International Nuclear Information System (INIS)

    Sredniawski, J.

    1978-01-01

    The vacuum system transient simulator (VSTS) models transient gas transport throughout complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. VSTS is capable of treating gas models of up to 10 species, for all flow regimes from pure molecular to continuum. Viscous interactions between species are considered as well as non-uniform temperature of a system. Although this program was specifically developed for use on the Tokamak Fusion Test Reactor (TFTR) project at Princeton, it is a generalized tool capable of handling a broad range of vacuum system problems. During the TFTR engineering design phase, VSTS has been used in many applications. Two applications selected for presentation are: (1) torus vacuum pumping system performance between 400 Ci tritium pulses and (2) tritium backstreaming to neutral beams during pulses

  14. Vacuum system transient simulator and its application to TFTR

    International Nuclear Information System (INIS)

    Sredniawski, J.

    1977-01-01

    The vacuum system transient simulator (VSTS) models transient gas transport throughout complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. VSTS is capable of treating gas models of up to 10 species, for all flow regimes from pure molecular to continuum. Viscous interactions between species are considered as well as non-uniform temperature of a system. Although this program was specifically developed for use on the Tokamak Fusion Test Reactor (TFTR) project at Princeton, it is a generalized tool capable of handling a broad range of vacuum system problems. During the TFTR engineering design phase, VSTS has been used in many applications. Two applications selected for presentation are: torus vacuum pumping system performance between 400 Ci tritium pulses and tritium backstreaming to neutral beams during pulses

  15. Cold Vacuum Drying Instrument Air System Design Description. System 12

    International Nuclear Information System (INIS)

    SHAPLEY, B.J.; TRAN, Y.S.

    2000-01-01

    This system design description (SDD) addresses the instrument air (IA) system of the spent nuclear fuel (SNF). This IA system provides instrument quality air to the Cold Vacuum Drying (CVD) Facility. The IA system is a general service system that supports the operation of the heating, ventilation, and air conditioning (HVAC) system, the process equipment skids, and process instruments in the CVD Facility. The following discussion is limited to the compressor, dryer, piping, and valving that provide the IA as shown in Drawings H-1-82222, Cold Vacuum Drying Facility Mechanical Utilities Compressed and Instrument Air PandID, and H-1.82161, Cold Vacuum Drying Facility Process Equipment Skid PandID MCO/Cusk Interface. Figure 1-1 shows the physical location of the 1A system in the CVD Facility

  16. 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

  17. Design and issues of the ITER in-vessel components: ITER Joint central team and home teams

    International Nuclear Information System (INIS)

    Parker, R.R.

    1998-01-01

    This paper surveys the status of the design of the in-vessel components for ITER, in particular the major components, namely the vacuum vessel, blanket and first wall, and divertor, and the interface of selected ancillary systems such as those used for RF heating and current drive, and for diagnostics. The vacuum vessel is a double-walled structure constructed from two toroidal shells joined by ribs. The space between the skins is filled with shield plates directly cooled by water. The structural material is 316 LN IG (ITER grade). Toroidal supports joining the vessel midplane ports with the TF structure limit possible differential toroidal displacements, as might occur due to seismic or vertical displacement events (VDEs). A variety of load conditions corresponding to normal and off-normal loads have been considered and in all cases peak vessel stresses are within allowables. The blanket system consists of approximately 700 modules, each weighing ∝4 t. The integrated first wall consists of a beryllium-tiled copper mat bonded to the water-cooled SS shield block. The copper mat functions as a heat sink and has imbedded in it an array of SS tubes providing water cooling. The modules are mechanically attached to a toroidal backplate. Loads due to centered disruptions are reacted via hoop stress in the backplate, whereas net vertical and horizontal loads such as those arising from VDEs are transferred through the backplate and divertor supports to the vessel. (orig.)

  18. ISR vacuum system

    CERN Multimedia

    CERN PhotoLab

    1970-01-01

    A pressure of 5 x 10-11 Torr has been obtained repreatedly in this pilot section of the ISR vacuum system. The pilot section is 45 m long is pumped by 9 sputter-ion pumps pf 350 l/s pumping speed, and is baked out at 200 degrees C before each pump down.

  19. Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

    International Nuclear Information System (INIS)

    Laurent Gargiulo, L.; Cordier, J.-J.; Samaille, F.; Grisolia, Ch.; Perrot, Y.; Olivier, D.; Friconneau, J.-P.; Palmer, J.

    2006-01-01

    The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier. This project called AIA (Articulated Inspection Arm) is currently being developed at CEA under a European EFDA work program. The paper describes the detailed design, the manufacturing processes and the results of the first module test campaign in the CEA Tore Supra ME60 facility, at representative vacuum, temperature and nominal loading conditions. The second part of this work that is reported in the paper, concerns the description of the whole integration of the device on the Tore Supra tokamak that is foreseen to be operated on Tore Supra early 2007. The deployer system and the 10 m long storage vacuum vessel are presented. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development is presented in the paper. It will allow close inspection of the Tore Supra Plasma Facing Components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. Such viewing process could be used on ITER during the early stage of operation under a limited radiation level. The AIA technology is also showing promising potential for generic application in alternative systems for ITER. The feasibility study for viewing inspection of the beam line components in the neutral beam test facility is presented. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be

  20. Design considerations for ITER magnet systems

    International Nuclear Information System (INIS)

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

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is now completing a definition phase as a beginning of a three-year design effort. Preliminary parameters for the superconducting magnet system have been established to guide further and more detailed design work. Radiation tolerance of the superconductors and insulators has been of prime importance, since it sets requirements for the neutron-shield dimension and sensitively influences reactor size. The major levels of mechanical stress in the structure appear in the cases of the inboard legs of the toroidal-field (TF) coils. The cases of the poloidal-field (PF) coils must be made thin or segmented to minimize eddy current heating during inductive plasma operation. As a result, the winding packs of both the TF and PF coils includes significant fractions of steel. The authors present here preliminary ITER magnet systems design parameters taken from trade studies, design, and analyses performed by the Home Teams of the four ITER participants, by the ITER Magnet Design Unit in Garching, and by other participants at workshops organized by the Magnet Design Unit

  1. Design and construction of vacuum control system on EAST

    International Nuclear Information System (INIS)

    Wang, L.; Zhang, Y.; Hu, Q.S.; Wang, X.M.; Zhang, X.D.; Hu, J.S.; Yang, Y.; Gu, X.M.

    2008-01-01

    The construction of experimental advanced superconducting tokamak (EAST) was finished at the end of 2006 in Hefei, China. Its vacuum system, an important subsystem, has been commissioned in February 2006. The design and construction of this vacuum control system are described in this paper. The requirements for remote automation, distributed control and centralized management, high reliability and expansibility have been taken into account in the design. There are three levels of control in vacuum control system. The bottom level control is performed on the local instruments manually; the medium level control is based on Siemens S7-400 PLC; the top level control is conducted on IPCs with communication through profi b us network. In addition remote handling and centralized monitoring could be realized by a remote control server. The control system could achieve pumping and fueling of the whole vacuum system. Besides that, it also includes the data acquisition of the pressure and temperature. The details are discussed on the monitoring of vacuum system states including cooling water, power and compressed air, etc., safeguards of plasma chamber and cryostat chamber and vacuum equipments, choosing of control modes corresponding to the plasma discharge and wall conditioning. At the end, the parts of EAST device protection system related to vacuum and gas injection system will also be introduced

  2. Overview of the ITER Tokamak complex building and integration of plant systems toward construction

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, Jean-Jacques, E-mail: jean-jacques.cordier@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bak, Joo-Shik [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Baudry, Alain [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Benchikhoune, Magali [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Carafa, Leontin; Chiocchio, Stefano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Darbour, Romaric [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Elbez, Joelle; Di Giuseppe, Giovanni; Iwata, Yasuhiro; Jeannoutot, Thomas; Kotamaki, Miikka; Kuehn, Ingo; Lee, Andreas; Levesy, Bruno; Orlandi, Sergio [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Packer, Rachel [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Patisson, Laurent; Reich, Jens; Rigoni, Giuliano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2015-10-15

    The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed

  3. The HIE-ISOLDE Vacuum System

    CERN Document Server

    Vandoni, G; Radwan, K; Chiggiato, P

    2014-01-01

    The High Intensity and Energy Isolde (HIE-Isolde) project aims at increasing the energy and intensity of the radioactive ion beams (RIB) delivered by the present Rex-Isolde facility. Energy up to 10MeV/amu will be reached by a new post-accelerating, superconducting (SC) linac. Beam will be delivered via a HEBT to three experimental stations for nuclear physics. To keep the SC linac compact and avoid cold-warm transitions, the cryomodules feature a common beam and insulation vacuum. Radioactive ion beams require a hermetically sealed vacuum, with transfer of the effluents to the nuclear ventilation chimney. Hermetically sealed, dry, gas transfer vacuum pumps are preferred to gas binding pumps, for an optimized management of radioactive contamination risk during maintenance and intervention. The vacuum system of the SC-linac is isolated by two fast valves, triggered by fast reacting cold cathode gauges installed on the warm linac, the HEBT and the experimental stations. Rough pumping is distributed, while the H...

  4. ITER safety task NID-5a: ITER tritium environmental source terms - safety analysis basis

    International Nuclear Information System (INIS)

    Natalizio, A.; Kalyanam, K.M.

    1994-09-01

    The Canadian Fusion Fuels Technology Project's (CFFTP) is part of the contribution to ITER task NID-5a, Initial Tritium Source Term. This safety analysis basis constitutes the first part of the work for establishing tritium source terms and is intended to solicit comments and obtain agreement. The analysis objective is to provide an early estimate of tritium environmental source terms for the events to be analyzed. Events that would result in the loss of tritium are: a Loss of Coolant Accident (LOCA), a vacuum vessel boundary breach. a torus exhaust line failure, a fuelling machine process boundary failure, a fuel processing system process boundary failure, a water detritiation system process boundary failure and an isotope separation system process boundary failure. 9 figs

  5. 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.

  6. PC driven integrated vacuum system

    International Nuclear Information System (INIS)

    Curuia, Marian; Culcer, Mihai; Brandea, Iulian; Anghel, Mihai

    2001-01-01

    The monitoring of industrial plants by virtual instrumentation represents the most modern trend in the domain of electronic equipment. The integrated vacuum system presented here has several facilities, including the automated data storing of measurement results on hard disk and providing warning messages for operators when the measured parameters are lower or higher upper than the fixed values. The system can also work stand-alone, receiving the commands from the keyboards placed on his front panel but, when it is included in a automation complex system, a remote control from PC is necessary . Both parts of the system, power supply unit for turbo-molecular pump and the vacuum gage, are controlled by an 80C31 microcontroller. Because this microcontroller has a built-in circuitry for a serial communication, we established a serial communication between the PC and the power supply unit for turbo-molecular pump and the vacuum gage, according to the RS-232 hardware standard. As software, after careful evaluation of several options, we chose to develop a hybrid software packing using two different software development tools: LabVIEW, and assembly language. We chose LabVIEW because it is dedicated to data acquisition and communications, containing libraries for data collection, analysis, display and storage. (authors)

  7. 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

  8. 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

  9. Progress in standardization for ITER Remote Handling control system

    International Nuclear Information System (INIS)

    Hamilton, David Thomas; Tesini, Alessandro; Ranz, Roberto; Kozaka, Hiroshi

    2014-01-01

    Graphical abstract: - Highlights: • Standard parts specified for ITER Remote Handling (RH) control system. • Standard approach for VR modeling of structural deformations in real-time. • RH Core System produced as standard platform for RH controller applications. • Synthetic Viewing investigated and demonstrated. • Structured language defined for RH operation procedures and motion sequences. - Abstract: An integrated control system architecture has been defined for the ITER Remote Handling (RH) equipment systems, and work has been continuing to develop and validate standards for this architecture. Evaluations of standard parts and a standard control room work-cell have contributed to an update of the RH Control System Design Handbook, while R and D activities have been carried out to validate concepts for standard solutions to ITER RH problems: the use of a standard master arm with different slave arms, the achievement of high accuracy tracking of RH operations within virtual reality, and condition monitoring of RH equipment systems. The standardization efforts have been consolidated through the development of a freely distributable software platform to support the adoption of the ITER RH standards. The RH Core System installs on top of the CODAC Core System and provides the basic platform for the development of ITER RH equipment controller applications. The standardization work has continued in the areas of RH viewing, network communication protocols, and a structured language for programming ITER RH operations. Prototyping has been done on high-level control system applications, and R and D has been carried out in the area of synthetic viewing for ITER RH. These developments will be reflected in a new version of the RH Core System to be produced during 2013

  10. Progress in standardization for ITER Remote Handling control system

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, David Thomas, E-mail: david.hamilton@iter.org [ITER Organization, Route de Vinon, 13115 St. Paul-lez-Durance (France); Tesini, Alessandro [ITER Organization, Route de Vinon, 13115 St. Paul-lez-Durance (France); Ranz, Roberto [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Kozaka, Hiroshi [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan)

    2014-10-15

    Graphical abstract: - Highlights: • Standard parts specified for ITER Remote Handling (RH) control system. • Standard approach for VR modeling of structural deformations in real-time. • RH Core System produced as standard platform for RH controller applications. • Synthetic Viewing investigated and demonstrated. • Structured language defined for RH operation procedures and motion sequences. - Abstract: An integrated control system architecture has been defined for the ITER Remote Handling (RH) equipment systems, and work has been continuing to develop and validate standards for this architecture. Evaluations of standard parts and a standard control room work-cell have contributed to an update of the RH Control System Design Handbook, while R and D activities have been carried out to validate concepts for standard solutions to ITER RH problems: the use of a standard master arm with different slave arms, the achievement of high accuracy tracking of RH operations within virtual reality, and condition monitoring of RH equipment systems. The standardization efforts have been consolidated through the development of a freely distributable software platform to support the adoption of the ITER RH standards. The RH Core System installs on top of the CODAC Core System and provides the basic platform for the development of ITER RH equipment controller applications. The standardization work has continued in the areas of RH viewing, network communication protocols, and a structured language for programming ITER RH operations. Prototyping has been done on high-level control system applications, and R and D has been carried out in the area of synthetic viewing for ITER RH. These developments will be reflected in a new version of the RH Core System to be produced during 2013.

  11. The vacuum interlock system for the CELSIUS ring

    International Nuclear Information System (INIS)

    Gajewski, K.

    1990-01-01

    A vacuum interlock system has been designed and built for the CELSIUS storage ring. The ultrahigh-vacuum system of CELSIUS has a design pressure of 10 -11 mbar. This is achieved by using vacuum-fired stainless-steel chambers, baking the whole ring to 300degC and running some 50 sputter ion and titanium sublimation pumps. The turbopumps, combined with roughing pumps, are used during the pump-down and the bake-out. The pressure is monitored by Penning vacuum gauges. There is a number of programmable pressure thresholds set to trigger various events (like closing the sector valves, disabling the bake-out, etc.). The interlock system is based on the Mitsubishi programmable logic controller (PLC). An IBM PC is used as an operator's console. The operation and performance of the system is described. On the basis of present experience an upgrading of the system is suggested. (orig.)

  12. ISR vacuum system

    CERN Multimedia

    CERN PhotoLab

    1971-01-01

    Some of the most important components of the vacuum system are shown. At the left, the rectangular box is a sputter-ion pump inside its bake-out oven. The assembly in the centre includes a sector valve, three roughing valves, a turbomolecular pump, a rotary backing pump and auxiliary equipment. At the right, the small elbow houses a Bayard-

  13. Effect of geometrical imperfection on buckling failure of ITER VVPSS tank

    International Nuclear Information System (INIS)

    Jha, Saroj Kumar; Gupta, Girish Kumar; Pandey, Manish Kumar; Bhattacharya, Avik; Jogi, Gaurav; Bhardwaj, Anil Kumar

    2015-01-01

    The 'Vacuum Vessel Pressure Suppression System' (VVPSS) is Part of ITER machine, which is designed to protect the ITER Vacuum Vessel and its connected systems, from an over-pressure situation. It is comprised of a partially evacuated tank of stainless steel approximately 46 meters long and 6 meters in diameter and thickness 30mm. It is to hold approximately 675 tonnes of water at room temperature to condense the steam resulting from the adverse water leakage into the Vacuum Vessel chamber. For any vacuum vessel, geometrical imperfection has significant effect on buckling failure and structural integrity. Major geometrical imperfection in VVPSS tank depends on form tolerances. To study the effect of geometrical imperfection on buckling failure of VVPSS tank, finite element analysis (FEA) has been performed in line with ASME section VIII division 2 part 5, 'design by analysis method'. Linear buckling analysis has been performed to get the buckled shape and displacement. Geometrical imperfection due to form tolerance is incorporated in FEA model of VVPSS tank by scaling the resulted buckled shape by a factor '60'. This buckled shape model is used as input geometry for plastic collapse and buckling failure assessment. Plastic collapse and buckling failure of VVPSS tank has been assessed by using the elastic-plastic analysis method. This analysis has been performed for different values of form tolerance. The results of analysis show that displacement and load proportionality factor (LPF) vary inversely with form tolerance. For higher values of form tolerance LPF reduces significantly with high values of displacement. (author)

  14. Bake-Out Mobile Controls for Large Vacuum Systems

    CERN Document Server

    Blanchard, S; Gomes, P; Pereira, H; Kopylov, L; Merker, S; Mikheev, M

    2014-01-01

    Large vacuum systems at CERN (Large Hadron Collider - LHC, Low Energy Ion Rings - LEIR...) require bake-out to achieve ultra-high vacuum specifications. The bake-out cycle is used to decrease the outgassing rate of the vacuum vessel and to activate the Non-Evaporable Getter (NEG) thin film. Bake-out control is a Proportional-Integral-Derivative (PID) regulation with complex recipes, interlocks and troubleshooting management and remote control. It is based on mobile Programmable Logic Controller (PLC) cabinets, fieldbus network and Supervisory Control and Data Acquisition (SCADA) application. The CERN vacuum installations include more than 7 km of baked vessels; using mobile cabinets reduces considerably the cost of the control system. The cabinets are installed close to the vacuum vessels during the time of the bake-out cycle. Mobile cabinets can be used in any of the CERN vacuum facilities. Remote control is provided through a fieldbus network and a SCADA application

  15. 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.

  16. 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.

  17. Development of a large lithium coolant system for operation under vacuum

    International Nuclear Information System (INIS)

    Kolowith, R.; Schwartz, K.E.; Meadows, G.E.; Berg, J.D.

    1983-11-01

    Argon and vacuum systems for the Experimental Lithium System (ELS) were tested to demonstrate vacuum-break capability, vacuum pumping performance, and vacuum sensor compatibility with a hostile liquid metal vapor/aerosol environment. Mechanical, diffusion and cryogenic vacuum pumps were evaluated. High-vacuum levels in the 10 -3 Pa range were achieved over a 270 0 C flowing lithium system. Ionization, thermal conductivity, capacitance manometer, and compound-type pressure sensors were evaluated to determine the effects of this potentially deleterious environment. Screening elbows were evaluated as pressure sensor protective devices. A dual-purpose vacuum-level/nitrogen partial-pressure sensor was evaluated as a means of detecting air in-leakage. Several types of static mechanical vacuum seals were also evaluated. Measurements of the vapor/aerosol generation were made at several system locations and operating conditions

  18. The vacuum interlock system for the PETRA III beamlines

    Energy Technology Data Exchange (ETDEWEB)

    Degenhardt, Markus; Hahn, Ulrich; Hesse, Mathias; Schulte-Schrepping, Horst [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany)

    2008-07-01

    The storage ring PETRA at DESY in Hamburg is being reconstructed into the third generation source for synchrotron radiation, PETRA III. The up to 100 m long beamlines are large UHV-systems that guide the synchrotron radiation from the storage ring to the experiments. Each beamline will be equipped with a vacuum interlock system to ensure the safe operation of the vacuum components. In particular the task of the vacuum interlock is to prevent faulty operations that can cause a ventilation of the vacuum system or a damage of vacuum components by the high power synchrotron radiation beam. The interlock system will be implemented as a PLC that is connected to a distributed input/output layer via a field bus system. As a specialty, the PLC will be realised as a soft-PLC running on a PC with a real time windows operating system. Another specialty is the visualisation and remote control of the vacuum interlock system by means of a website. At the beamline the interlock will be operated via a touch panel that displays the visualisation website. Additionally, the interlock can be remotely operated from any location by opening the visualisation website with a browser. The interlock is protected against unauthorised operation by a login page. All relevant interlock data will be fed into the existing network-based archive system.

  19. Designing a prototype of the ITER pulse scheduling system

    International Nuclear Information System (INIS)

    Yamamoto, T.; Yonekawa, I.; Ohta, K.; Hosoyama, H.; Hashimoto, Y.; Wallander, A.; Winter, A.; Sugie, T.; Kusama, Y.; Kawano, Y.; Yoshino, R.

    2012-01-01

    Highlights: ► We designed a prototype of the ITER pulse scheduling system. ► Structure of ITER pulse schedules was designed. ► Validation and automatic value assignment functions were adopted. ► A prototype will be implemented in 2011. - Abstract: A prototype of the ITER pulse scheduling system that prepares and manages parameters for ITER plasma operations has been designed. Based on the analyzed requirements on the system, structure of the parameters and necessary functions were determined. Segment and module structures were tuned to the ITER requirements. Three types of validations assure sanity of the parameters. The design limits check and the operation window check verify whether the values of the parameters do not exceed the limits. The consistency check calculates dependency among parameters in accordance with logics described in a scripting language. The ITER pulse scheduling system provides interface with a physics model and simulator. Some abstract physics parameters are converted to engineering parameters with the physics simulation. The results of simulation such as plasma characteristics of specified parameters are also shown to the researchers. The tool to specify the parameters is data-driven. Therefore, it is flexible for changes of number of the parameters. A prototype is being implemented in 2011. Using the prototype, this design will be verified and refined. The evaluation of the prototype will be a basis of the final production of the ITER pulse scheduling system.

  20. Development of ITER CODAC compatible gyrotron local control system and its operation

    International Nuclear Information System (INIS)

    Ohshima, Katsumi; Oda, Yasuhisa; Takahashi, Koji; Terakado, Masayuki; Ikeda, Ryosuke; Moriyama, Shinichi; Kajiwara, Ken; Sakamoto, Keishi; Hayashi, Kazuo

    2016-03-01

    In Japan Atomic Energy Agency, an ITER relevant control system for ITER gyrotron was developed according to Plant Control Design Handbook. This control system was developed based on ITER CODAC Core System and implemented state machine control of gyrotron operation system, sequential timing control of gyrotron oscillation startup, and data acquisition. The operation of ITER 170 GHz gyrotron was demonstrated with ITER relevant power supply configuration. This system is utilized for gyrotron operation test for ITER procurement. This report describes the architecture of gyrotron local control system, its basic and detailed design, and recent operation results. (author)

  1. Structural analysis of ITER sub-assembly tools

    International Nuclear Information System (INIS)

    Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Lee, J.H.; Kim, K.K.; Im, K.; Shaw, R.

    2011-01-01

    The ITER Tokamak assembly tools are purpose-built assembly tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. The sector sub-assembly tools descried in this paper are main assembly tools to assemble vacuum vessel, thermal shield and toroidal filed coils into a complete 40 o sector. The 40 o sector sub-assembly tools are composed of sector sub-assembly tool, including radial beam, vacuum vessel supports and mid-plane brace tools. These tools shall have sufficient strength to transport and handle heavy weight of the ITER Tokamak machine reached several hundred tons. Therefore these tools should be designed and analyzed to confirm both the strength and structural stability even in the case of conservative assumptions. To verify structural stabilities of the sector sub-assembly tools in terms of strength and deflection, ANSYS code was used for linear static analysis. The results of the analysis show that these tools are designed with sufficient strength and stiffness. The conceptual designs of these tools are briefly described in this paper also.

  2. 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.

  3. 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

  4. Experimental validation of a method for performance monitoring of the impurity processing stage in the TEP system of ITER

    International Nuclear Information System (INIS)

    Bornschein, B.; Corneli, D.; Glugla, M.; Guenther, K.; Le, T.L.; Simon, K.H.

    2007-01-01

    The Tokamak Exhaust Processing (TEP) system within the Tritium Plant of ITER needs to be designed such that tritium is recovered from all exhaust gases produced during different modes and operational conditions of the vacuum vessel. The reference process for the TEP system of ITER is called CAPER and comprises three different, consecutive steps to recover hydrogen isotopes at highest purity for direct transfer to the cryogenic Isotope Separation system. The second step ('impurity processing', IP) is carried out in a closed loop involving heterogeneously catalyzed cracking or conversion reactions to liberate tritium from tritiated hydrocarbons or tritiated water combined with permeation of hydrogen isotopes through a Pd/Ag permeator. This combination shifts chemical equilibria towards dehydrogenation and, therefore, enables detritiation factors higher than 1000 in the IP stage. Such a high decontamination factor requires the optimal performance of the permeator, which on the other hand is operated under conditions which provoke coking of the permeator membrane by hydrocarbon cracking. For this reason the permeator in the impurity processing loop needs to be repeatedly regenerated in order to sustain decontamination factors higher/in the order of 1000. At the Tritium Laboratory Karlsruhe (TLK) a method to measure the actual performance of the second stage of the CAPER process has been developed. This method has been successfully tested with the CAPER facility and appears feasible for the TEP system of ITER

  5. ITER vacuum vessel design and electromagnetic analysis on in-vessel components

    International Nuclear Information System (INIS)

    Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.; Iizuka, T.

    1995-01-01

    Major functional requirements for the vacuum vessel are to provide the first safety barrier and to support electromagnetic loads due to plasma disruptions and vertical displacement events, and to withstand plausible accidents without losing confinement. A double wall structure concept has been developed for the vacuum vessel due to its beneficial characteristics from the viewpoints of structural integrity and electrical continuity. An electromagnetic analysis of the blanket modules and the vacuum vessel has been performed to investigate force distributions on in-vessel components. According to the vertical displacement events (VDE) scenario, which assumes a critical q-value of 1.5, the total downward vertical force, induced by coupling between the eddy current and external fields, is about 110 MN. We have performed a stress analysis for the vacuum vessel using the VDE disruption forces acting on the blankets, and a maximum stress intensity of 112 MPa was obtained in the vicinity of the lower support of the vessel. (orig.)

  6. Development of simulator for remote handling system of ITER blanket

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakanhira, Masataka; Matsumoto, Yasuhiro; Shibanuma, K.

    2007-01-01

    The maintenance activity in the ITER has to be performed remotely because 14 MeV neutron caused by fusion reaction induces activation of structural material and emission of gamma ray. In general, it is one of the most critical issues to avoid collision between the remote maintenance system and in-vessel components. Therefore, the visual information in the vacuum vessel is required strongly to understand arrangement of these devices and components. However, there is a limitation of arrangement of viewing cameras in the vessel because of high intensity of gamma ray. It is expected that enough numbers of cameras and lights are not available because of arrangement restriction. Furthermore, visibility of the interested area such as the contacting part is frequently disturbed by the devices and components, thus it is difficult to recognize relative position between the devices and components only by visual information even if enough cameras and lights are equipped. From these reasons, the simulator to recognize the positions of each devices and components is indispensable for remote handling systems in fusion reactors. The authors have been developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robot simulation software ''ENVISION''. The simulator is connected to the control system of the manipulator which was developed as a part of the blanket maintenance system in the EDA and can reconstruct the positions of the manipulator and the blanket module using the position data of the motors through the LAN. In addition, it can provide virtual visual information, such as the connecting operation behind the blanket module with making the module transparent on the screen. It can be used also for checking the maintenance sequence before the actual operation. The developed simulator will be modified further adding other necessary functions and finally completed as a prototype of the actual simulator for the blanket remote handling system

  7. A description of the ITER's gas injection systems and current R and D activities

    International Nuclear Information System (INIS)

    Li, W.; Li, B.; Maruyama, S.; Jiang, T.; Yang, Y.; Xia, Z.W.; Zhang, Y.X.; Lu, J.

    2012-01-01

    The gas injection system (GIS) is an indispensable part of ITER fueling system. It deliveries the necessary gas species from tritium plant to vacuum vessel, pellet injection system or neutral beam for plasma operation and fusion power shutdown. In this paper, the current design status of GIS, including the previous design changes, is briefly described. As the GIS design justification and support, the experimental study on GIS response time is illustrated. The factors delayed the GIS response time are identified, and two kinds of control mode are proved to be effective for improving the GIS response time. The exploration on magnetic shield design shows the discrepancy of shielding performance occurs in the case of the paralleling external magnetic field to the sample cylinder. These R and D works prove the design feasibility in some ways, and support possible solutions for design challenges as alternative design options.

  8. Continuous vacuum processing system for quartz crystal resonators

    International Nuclear Information System (INIS)

    Ney, R.J.; Hafner, E.

    1979-01-01

    An ultrahigh vacuum continuous cycle quartz crystal fabrication facility has been developed that assures an essentially contamination-free environment throughout the final manufacturing steps of the crystal unit. The system consists of five essentially tubular vacuum chambers that are interconnected through gate valves. The unplated crystal resonators, mounted in ceramic flatback frames and loaded on carrier trays, enter the vacuum system through an entrance air lock, are UV/ozone cleaned, baked at 300 0 C, plated to frequency, thermocompression sealed, and exit as completed crystal units through an exit air lock, while the bake, plate and seal chambers remain under continuous vacuum permanently. In-line conveyor belts are used, in conjunction with balanced vacuum manipulators, to move the resonator components to the various work stations. Unique high density, highly directional nozzle beam evaporation sources, capable of long term operation without reloading, are used for electroding the resonators simultaneously on both sides. The design goal for the system is a production rate of 200 units per 8 hour day; it is adaptable to automatic operation

  9. 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.)

  10. Internal motion in high vacuum systems

    Science.gov (United States)

    Frank, J. M.

    Three transfer and positioning mechanisms have been developed for the non-air exposed, multistep processing of components in vacuum chambers. The functions to be performed in all of the systems include ultraviolet/ozone cleaning, vacuum baking, deposition of thin films, and thermocompression sealing of the enclosures. Precise positioning of the components is required during the evaporation and sealing processes. The three methods of transporting and positioning the components were developed to accommodate the design criteria and goals of each individual system. The design philosophy, goals, and operation of the three mechanisms are discussed.

  11. Conceptual design finalisation of the ITER In-Vessel Viewing and Metrology System (IVVS)

    Energy Technology Data Exchange (ETDEWEB)

    Dubus, Gregory, E-mail: gregory.dubus@f4e.europa.eu [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Puiu, Adrian; Damiani, Carlo; Van Uffelen, Marco; Lo Bue, Alessandro; Izquierdo, Jesus; Semeraro, Luigi [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Martins, Jean-Pierre; Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components. Periodically or on request, the IVVS probes will be deployed into the Vacuum Vessel from their storage positions (still within the ITER primary confinement) in order to perform both viewing and metrology on plasma facing components (blanket, divertor, heating/diagnostic plugs, test blanket modules) and, more generically, to provide information on the status of the in-vessel components. In 2011, the IO proposed to simplify and strengthen the six IVVS port extensions situated at the divertor level. Among other important consequences, such as the relocation of the Glow Discharge Cleaning (GDC) electrodes at other levels of the machine, this major design change implied the need for a substantial redesign of the IVVS plug, which took part to an on-going effort to bring the integrated IVVS concept – including the scanning probe and its deployment system – to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the various design and R and D activities in progress: plug design integration, probe concept validation under environmental conditions, development of a metrology strategy, the whole supported by a nuclear analysis.

  12. The New Control System for the Vacuum of ISOLDE

    CERN Document Server

    Blanchard, S; Bernard, FB; Blanco, E; Gomes, P; Vestergard, H; Willeman, D

    2011-01-01

    The On-Line Isotope Mass Separator (ISOLDE) is a facility dedicated to the production of radioactive ion beams for nuclear and atomic physics. From ISOLDE vacuum sectors to the pressurized exhaust gas storage tanks there are up to five stages of pumping for a total of more than one hundred pumps including turbo-molecular, cryogenic, dry, membrane and oil pumps. The ISOLDE vacuum control system is critical; the volatile radioactive elements present in the exhaust gases and the high and ultra high vacuum pressure specifications require a complex control and interlock system. This paper describes the reengineering of the control system developed using the CERN UNICOS-CPC framework. An additional challenge has been the usage of the UNICOS-CPC in a vacuum domain for the first time. The process automation provides multiple operating modes (rough pumping, bake-out, high vacuum pumping, regeneration for cryo-pumped sectors, venting, etc). The control system is composed of local controllers driven by...

  13. Evaluation of CBA first string full cell vacuum system

    International Nuclear Information System (INIS)

    Foerster, C.L.; Briggs, J.; Christianson, C.; Stattel, P.

    1983-01-01

    The CBA (Colliding Beam Accelerator, formerly known as ISABELLE) Full Cell Magnet System consisting of six superconducting dipole magnets and two superconducting quadrupole magnets requires two separate vacuum systems. One, known as beam vacuum operates below 3 x 10 -11 Torr and the other, known as insulating vacuum, operates at less than 10 -7 Torr to isolate cryo circuits from atmosphere and from the uhv beam tubes. The uhv bore tube is isolated from the 4.0 0 K magnet by thirty-six (36) layers of superinsulation and insulating vacuum. Heat load measurements on the bore tube have been completed and found to agree with data obtained in smaller controlled experiments. Measurements of helium, accumulated on cryogenic pumped charcoal panels over many weeks, have verified sensitive helium mass spectrometer leak detection methods for vacuum integrity, providing sound design of the welded complex. The Full Cell was assembled and operated under conditions that would exist in the completed machine. Pressures below 2 x 10 -11 Torr beam vacuum requirement and below 2 x 10 -7 Torr insulating vacuum, were routinely achieved during all phases of the Full Cell operation and support systems testing

  14. Thermal Vacuum Integrated System Test at B-2

    Science.gov (United States)

    Kudlac, Maureen T.; Weaver, Harold F.; Cmar, Mark D.

    2012-01-01

    The National Aeronautics and Space Administration (NASA) Glenn Research Center (GRC) Plum Brook Station (PBS) Space Propulsion Research Facility, commonly referred to as B-2, is NASA s third largest thermal vacuum facility. It is the largest designed to store and transfer large quantities of liquid hydrogen and liquid oxygen, and is perfectly suited to support developmental testing of chemical propulsion systems as well as fully integrated stages. The facility is also capable of providing thermal-vacuum simulation services to support testing of large lightweight structures, Cryogenic Fluid Management (CFM) systems, electric propulsion test programs, and other In-Space propulsion programs. A recently completed integrated system test demonstrated the refurbished thermal vacuum capabilities of the facility. The test used the modernized data acquisition and control system to monitor the facility during pump down of the vacuum chamber, operation of the liquid nitrogen heat sink (or cold wall) and the infrared lamp array. A vacuum level of 1.3x10(exp -4)Pa (1x10(exp -6)torr) was achieved. The heat sink provided a uniform temperature environment of approximately 77 K (140deg R) along the entire inner surface of the vacuum chamber. The recently rebuilt and modernized infrared lamp array produced a nominal heat flux of 1.4 kW/sq m at a chamber diameter of 6.7 m (22 ft) and along 11 m (36 ft) of the chamber s cylindrical vertical interior. With the lamp array and heat sink operating simultaneously, the thermal systems produced a heat flux pattern simulating radiation to space on one surface and solar exposure on the other surface. The data acquired matched pretest predictions and demonstrated system functionality.

  15. Arc detection for the ICRF system on ITER

    Science.gov (United States)

    D'Inca, R.

    2011-12-01

    The ICRF system for ITER is designed to respect the high voltage breakdown limits. However arcs can still statistically happen and must be quickly detected and suppressed by shutting the RF power down. For the conception of a reliable and efficient detector, the analysis of the mechanism of arcs is necessary to find their unique signature. Numerous systems have been conceived to address the issues of arc detection. VSWR-based detectors, RF noise detectors, sound detectors, optical detectors, S-matrix based detectors. Until now, none of them has succeeded in demonstrating the fulfillment of all requirements and the studies for ITER now follow three directions: improvement of the existing concepts to fix their flaws, development of new theoretically fully compliant detectors (like the GUIDAR) and combination of several detectors to benefit from the advantages of each of them. Together with the physical and engineering challenges, the development of an arc detection system for ITER raises methodological concerns to extrapolate the results from basic experiments and present machines to the ITER scale ICRF system and to conduct a relevant risk analysis.

  16. Operating status of TARN vacuum system and future tasks

    International Nuclear Information System (INIS)

    Chida, Katsuhisa; Tsujikawa, Hiroshi; Mizobuchi, Akira

    1981-01-01

    TARN (Test Accumulation Ring for Numatron) was constructed for the purpose of obtaining the fundamental data for high energy heavy ion accelerator (Numatron) project, which accelerates heavy ions up to uranium to 1 GeV/nucleon. Its vacuum is required to be 1 x 10 - 10 Torr or less on beam. In February, 1972, only the vacuum system was temporarily assembled, and the vacuum of 2 x 10 - 11 Torr was realized by baking at 300 deg C alone. In July, 1972, the assembling of the vacuum chamber into magnets was completed, and several test experiments were performed using the H 2+ beam from the SF cyclotron. In this report, first, the outline of the vacuum system, and next, its operation are described. For the reason of the purpose of the ring, the vacuum system is required to be atmospheric pressure to attach beam monitors and other measuring instruments just before the machine time. Therefore, it is an important task to make the evacuation time as short as possible. As future tasks, the examination on the material and shape of the chamber, the investigation of pump system (appropriate combination of ion pump, titanium sublimation pump, cryo-pump, molecular pump, etc.), the study on the measuring and control systems (accurate measurement of total pressure and partial pressure and the feedback to the protecting system), the studies of problems on the vacuum wall surface (surface treatment prior to assembling the chamber into the ring and the methods and the effects of baking and electric discharge cleaning) are included. (Wakatsuki, Y.)

  17. An integrated architecture for the ITER RH control system

    International Nuclear Information System (INIS)

    Hamilton, David Thomas; Tesini, Alessandro

    2012-01-01

    Highlights: ► Control system architecture integrating ITER remote handling equipment systems. ► Standard control system architecture for remote handling equipment systems. ► Research and development activities to validate control system architecture. ► Standardization studies to select standard parts for control system architecture. - Abstract: The ITER remote handling (RH) system has been divided into 7 major equipment system procurements that deliver complete systems (operator interfaces, equipment controllers, and equipment) according to task oriented functional specifications. Each equipment system itself is an assembly of transporters, power manipulators, telemanipulators, vehicular systems, cameras, and tooling with a need for controllers and operator interfaces. From an operational perspective, the ITER RH systems are bound together by common control rooms, operations team, and maintenance team; and will need to achieve, to a varying degree, synchronization of operations, co-operation on tasks, hand-over of components, and sharing of data and resources. The separately procured RH systems must, therefore, be integrated to form a unified RH system for operation from the RH control rooms. The RH system will contain a heterogeneous mix of specially developed RH systems and off-the-shelf RH equipment and parts. The ITER Organization approach is to define a control system architecture that supports interoperable heterogeneous modules, and to specify a standard set of modules for each system to implement within this architecture. Compatibility with standard parts for selected modules is required to limit the complexity for operations and maintenance. A key requirement for integrating the control system modules is interoperability, and no module should have dependencies on the implementation details of other modules. The RH system is one of the ITER Plant systems that are integrated and coordinated through the hierarchical structure of the ITER CODAC system

  18. ITER nuclear components, preparing for the construction and R&D results

    Science.gov (United States)

    Ioki, K.; Akiba, M.; Barabaschi, P.; Barabash, V.; Chiocchio, S.; Daenner, W.; Elio, F.; Enoeda, M.; Ezato, K.; Federici, G.; Gervash, A.; Grebennikov, D.; Jones, L.; Kajiura, S.; Krylov, V.; Kuroda, T.; Lorenzetto, P.; Maruyama, S.; Merola, M.; Miki, N.; Morimoto, M.; Nakahira, M.; Ohmori, J.; Onozuka, M.; Rozov, V.; Sato, K.; Strebkov, Yu; Suzuki, S.; Tanchuk, V.; Tivey, R.; Utin, Yu

    2004-08-01

    Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R&D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20° or 30°, on flow distribution tests of a two-channel model, on fabrication and testing of FW mock-ups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

  19. Efficient approach to simulate EM loads on massive structures in ITER machine

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Andreeva, Z.; Belov, A.; Belyakov, V.; Filatov, O. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Gribov, Yu.; Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Kukhtin, V.; Labusov, A.; Lamzin, E.; Lyublin, B.; Malkov, A.; Mazul, I. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Rozov, V.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Sychevsky, S., E-mail: sytch@sintez.niiefa.spb.su [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation)

    2013-10-15

    Highlights: ► A modelling technique to predict EM loads in ITER conducting structures is presented. ► The technique provides low computational cost and parallel computations. ► Detailed models were built for the systemvacuum vessel, cryostat, thermal shields”. ► EM loads on massive in-vessel structures were simulated with the use of local models. ► A flexible combination of models enables desired accuracy of load distributions. -- Abstract: Operation of the ITER machine is associated with high electromagnetic (EM) loads. An essential contributor to EM loads is eddy currents induced in passive conductive structures. Reasoning from the ITER construction, a modelling technique has been developed and applied in computations to efficiently predict anticipated loads. The technique allows us to avoid building a global 3D finite-element (FE) model that requires meshing of the conducting structures and their vacuum environment into 3D solid elements that leads to high computational cost. The key features of the proposed technique are: (i) the use of an existing shell model for the systemvacuum vessel (VV), cryostat, and thermal shields (TS)” implementing the magnetic shell approach. A solution is obtained in terms of a single-component, in this case, vector electric potential taken within the conducting shells of the “VV + cryostat + TS” system. (ii) EM loads on in-vessel conducting structures are simulated with the use of local FE models. The local models use either the 3D solid body or shell approximations. Reasoning from the simulation efficiency, the local boundary conditions are put with respect to the total field or an external field. The use of an integral-differential formulation and special procedures ensures smooth and accurate simulated distributions of fields from current sources of any geometry. The local FE models have been developed and applied for EM analyses of a variety of the ITER components including the diagnostic systems

  20. RAMI analysis of the ITER Central Safety System

    Energy Technology Data Exchange (ETDEWEB)

    Kitazawa, Sin-iti, E-mail: kitazawa.siniti@jaea.go.jp [ITER Project Unit, Japan Atomic Energy Agency (JAEA), Naka, 311-0193 Ibaraki (Japan); Okayama, Katsumi [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Neyatani, Yuzuru [ITER Project Unit, Japan Atomic Energy Agency (JAEA), Naka, 311-0193 Ibaraki (Japan); Sagot, Francois; Houtte, Didier van [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-06-15

    Highlights: • We performed the functional analysis of the ITER CSS. • We performed a failure mode analysis of the ITER CSS. • We estimated the reliability and availability of the ITER CSS. • The ITER RAMI approach was applied to the ITER CSS for technical risk control in the design phase. - Abstract: ITER is the first worldwide international project aiming to design a facility to produce nuclear fusion energy. The technical requirements of its plant systems have been established in the ITER Project Baseline. In the project, the Reliability, Availability, Maintainability and Inspectability (RAMI) approach has been adopted for technical risk control to help aid the design of the components in preparation for operation and maintenance. A RAMI analysis was performed on the conceptual design of the ITER Central Safety System (CSS). A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 2 main functions and 20 sub-functions. These functions were described using the IDEF0 method. Reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. Initial and expected scenarios were analyzed to define risk-mitigation actions. The inherent availability of the ITER CSS expected after implementation of mitigation actions was calculated to be 99.80% over 2 years, which is the typical interval of the scheduled maintenance cycles. This is consistent with the project required value of 99.9 ± 0.1%. A Failure Modes, Effects and Criticality Analysis 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 of the plasma operation. This analysis defined when risk mitigation actions were required in terms of design, testing, operation procedures and/or maintenance to reduce the risk levels and increase the availability of the

  1. Status of the ITER Ion Cyclotron H and CD system

    Energy Technology Data Exchange (ETDEWEB)

    Lamalle, P., E-mail: philippe.lamalle@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Beaumont, B.; Kazarian, F.; Gassmann, T. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Agarici, G. [Fusion for Energy, Carrer Josep Pla 2, Torres Diagonal Litoral Edificio B3, 08019 Barcelona (Spain); Ajesh, P. [ITER India, Institute for Plasma Research, Bhat, Gandhinagar 382424, Gujarat (India); Alonzo, T. [Solution F, Allée du Verdon, 13770 Venelles (France); Arambhadiya, B. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Argouarch, A. [CEA Cadarache, IRFM, F-13108 St-Paul-lez-Durance (France); Bamber, R. [EURATOM/CCFE Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Berger-By, G.; Bernard, J.-M.; Brun, C. [CEA Cadarache, IRFM, F-13108 St-Paul-lez-Durance (France); Carpentier, S. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Clairet, F.; Colas, L.; Courtois, X. [CEA Cadarache, IRFM, F-13108 St-Paul-lez-Durance (France); Davis, A. [EURATOM/CCFE Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Dechelle, C.; Doceul, L. [CEA Cadarache, IRFM, F-13108 St-Paul-lez-Durance (France); and others

    2013-10-15

    Highlights: ► We summarize the progress and outstanding issues in the development of the ITER Ion Cyclotron Heating and Current Drive (IC H and CD) system. ► The system is designed to robustly couple 20 MW in quasi-CW operation for a broad range of plasma scenarios, and is upgradeable to up to 40 MW. ► The design is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject. ► R and D is ongoing to validate key antenna components, and to qualify the radio-frequency (RF) sources and the transmission and matching components. ► Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. -- Abstract: The ongoing design of the ITER Ion Cyclotron Heating and Current Drive system (20 MW, 40–55 MHz) is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject, several of which are conflicting and/or still in a high state of flux. These requirements include operation over a broad range of plasma scenarios and magnetic fields (which prompts usage of wide-band phased antenna arrays), high radio-frequency (RF) power density at the first wall (and associated operation close to voltage and current limits), resilience to ELM-induced load variations, intense thermal and mechanical loads, long pulse operation, high system availability, efficient nuclear shielding, high density of antenna services, remote-handling ability, tight installation tolerances, and nuclear safety function as tritium confinement barrier. R and D activities are ongoing or in preparation to validate critical antenna components (plasma-facing Faraday screen, RF sliding contacts, RF vacuum windows), as well as to qualify the RF power sources and the transmission and matching components. Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. The paper

  2. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

  3. Progress in the design and R and D of the ITER In-Vessel Viewing and Metrology System (IVVS)

    Energy Technology Data Exchange (ETDEWEB)

    Dubus, Gregory, E-mail: gregory.dubus@f4e.europa.eu [Fusion for Energy, c/ Josep Pla, n°2 – Torres Diagonal Litoral – Edificio B3, 08019 Barcelona (Spain); Puiu, Adrian; Bates, Philip; Damiani, Carlo [Fusion for Energy, c/ Josep Pla, n°2 – Torres Diagonal Litoral – Edificio B3, 08019 Barcelona (Spain); Reichle, Roger; Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components, which in turn is related to the amount of mobilised dust present in the Vacuum Vessel. Periodically or on request, the IVVS scanning probes will be deployed into the Vacuum Vessel in order to acquire both visual and metrological data on plasma facing components (blanket, divertor, heating/diagnostic plugs, and test blanket modules). Recent design changes made to the six IVVS port extensions implied the need for a substantial redesign of the IVVS integrated concept – including the scanning probe and its deployment system – in order to bring it to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the concept design for IVVS as well as of the various engineering analyses and R and D activities carried out in support to this design: neutronic, seismic and electromagnetic analyses, probe actuation validation under environmental conditions.

  4. PC driven integrated vacuum system

    International Nuclear Information System (INIS)

    Curuia, M.; Culcer, M.; Brandea, I.; Anghel, M.

    2001-01-01

    The paper presents a integrated vacuum system which was designed and manufactured in our institute. The main parts of this system are the power supply unit for turbo-melecular pumps and the vacuummeter. Both parts of the system are driven by means of a personal computer using a serial communication, according to the RS 232 hardware standard.(author)

  5. Loads specification and embedded plate definition for the ITER cryoline system

    Science.gov (United States)

    Badgujar, S.; Benkheira, L.; Chalifour, M.; Forgeas, A.; Shah, N.; Vaghela, H.; Sarkar, B.

    2015-12-01

    ITER cryolines (CLs) are complex network of vacuum-insulated multi and single process pipe lines, distributed in three different areas at ITER site. The CLs will support different operating loads during the machine life-time; either considered as nominal, occasional or exceptional. The major loads, which form the design basis are inertial, pressure, temperature, assembly, magnetic, snow, wind, enforced relative displacement and are put together in loads specification. Based on the defined load combinations, conceptual estimation of reaction loads have been carried out for the lines located inside the Tokamak building. Adequate numbers of embedded plates (EPs) per line have been defined and integrated in the building design. The finalization of building EPs to support the lines, before the detailed design, is one of the major design challenges as the usual logic of the design may alter. At the ITER project level, it was important to finalize EPs to allow adequate design and timely availability of the Tokamak building. The paper describes the single loads, load combinations considered in load specification and the approach for conceptual load estimation and selection of EPs for Toroidal Field (TF) Cryoline as an example by converting the load combinations in two main load categories; pressure and seismic.

  6. Final Report on ITER Task Agreement 81-08

    Energy Technology Data Exchange (ETDEWEB)

    Richard L. Moore

    2008-03-01

    As part of an ITER Implementing Task Agreement (ITA) between the ITER US Participant Team (PT) and the ITER International Team (IT), the INL Fusion Safety Program was tasked to provide the ITER IT with upgrades to the fusion version of the MELCOR 1.8.5 code including a beryllium dust oxidation model. The purpose of this model is to allow the ITER IT to investigate hydrogen production from beryllium dust layers on hot surfaces inside the ITER vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). Also included in the ITER ITA was a task to construct a RELAP5/ATHENA model of the ITER divertor cooling loop to model the draining of the loop during a large ex-vessel pipe break followed by an in-vessel divertor break and compare the results to a simular MELCOR model developed by the ITER IT. This report, which is the final report for this agreement, documents the completion of the work scope under this ITER TA, designated as TA 81-08.

  7. 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.

  8. Safety characteristics of options for plasma-facing components for ITER and beyond

    International Nuclear Information System (INIS)

    Piet, S.J.; McCarthy, K.A.; Holland, D.F.; Longhurst, G.R.; Merrill, B.J.

    1991-01-01

    Plasma-facing components (PFC) likely dominate the safety hazards of the International Thermonuclear Experimental Reactor (ITER) and post-ITER machines. To gain regulatory approval and for fusion energy to fulfill its ultimate attractive safety and environmental potential, safety must be considered when selecting among PFC options. This paper summarizes current PFC safety information. PFC safety issues fall into seven areas: disruption tolerance, disruption severity, tritium inventory and permeation, accidental energy release, activation/toxin hazards, cooling disturbances, and system issues. RFC options include current ITER mainline options (Be or W coating, C tiles), variants on current ITER options, and liquid metal (LM) divertors. No PFC option that we have examined is free of critical safety concerns. There are also innovative ideas that may improve any PFC's performance -- super-permeable vacuum ducts, helium self-pumping, and gaseous divertors. We conclude with recommendations and a future strategy. 17 refs., 1 fig., 3 tabs

  9. Evaluation of the cryogenic mechanical properties of the insulation material for ITER Feeder superconducting joint

    Science.gov (United States)

    Wu, Zhixiong; Huang, Rongjin; Huang, ChuanJun; Yang, Yanfang; Huang, Xiongyi; Li, Laifeng

    2017-12-01

    The Glass-fiber reinforced plastic (GFRP) fabricated by the vacuum bag process was selected as the high voltage electrical insulation and mechanical support for the superconducting joints and the current leads for the ITER Feeder system. To evaluate the cryogenic mechanical properties of the GFRP, the mechanical properties such as the short beam strength (SBS), the tensile strength and the fatigue fracture strength after 30,000 cycles, were measured at 77K in this study. The results demonstrated that the GFRP met the design requirements of ITER.

  10. ITER assembly and maintenance

    International Nuclear Information System (INIS)

    Honda, T.; Davis, F.; Lousteau, D.

    1991-01-01

    This document is intended to describe the work conducted by the ITER Assembly and Maintenance (A and M) Design Unit and the supporting home teams during the ITER Conceptual Design Activities, carried out from 1988 through 1990. Its content consists of two main sections, i.e., Chapter III, which describes the identified tasks to be performed by the A and M system and a general description of the required equipment; and Chapter IV, which provides a more detailed description of the equipment proposed to perform the assigned tasks. A two-stage R and D program is now planned, i.e., (1) a prototype equipment functional tests using full scale mock-ups and (2) a full scale integration demonstration test facility with real components (vacuum vessel with ports, blanket modules, divertor modules, armor tiles, etc.). Crucial in-vessel and ex-vessel operations and the associated remote handling equipment, including handling of divertor plates and blanket modules will be demonstrated in the first phase, whereby the database needed to proceed with the engineering phase will be acquired. The second phase will demonstrate the ability of the overall system to execute the required maintenance procedures and evaluate the performance of the prototype equipment

  11. In situ conditioning for proton storage ring vacuum systems

    International Nuclear Information System (INIS)

    Blechschmidt, D.

    1978-01-01

    Average pressure and vacuum-stability limit as expected in the presence of a proton beam were measured after in situ treatments such as bakeout under various conditions, argon glow-discharge cleaning and sputter deposition of titanium. Measurements were carried out for test pipes made of stainless steel (untreated, electropolished, or cooled to 77 K), pure titanium and aluminum alloy. The measurement method used to obtain the vacuum-stability limit in the laboratory and in a prototype system is described. The results can be applied also to other systems of different geometry by use of scaling laws. In situ conditioning generally has a stronger influence on vacuum performance than a particular choice of material. Bakeout gives low average pressures and rather good vacuum stability. Glow discharges also increase the vacuum stability but have only a small effect on the static pressure. Coating the beam-pipe wall with titanium by in situ sputtering provides large linear pumping, thus a lower pressure and an extremely good vacuum stability

  12. Sewerage force adjustment technology for energy conservation in vacuum sanitation systems

    Science.gov (United States)

    Guo, Zhonghua; Li, Xiaoning; Kagawa, Toshiharu

    2013-03-01

    The vacuum sanitation is the safe and sound disposal approach of human excreta under the specific environments like flights, high speed trains and submarines. However, the propulsive force of current systems is not adjustable and the energy consumption does not adapt to the real time sewerage requirement. Therefore, it is important to study the sewerage force adjustment to improve the energy efficiency. This paper proposes an energy conservation design in vacuum sanitation systems with pneumatic ejector circuits. The sewerage force is controlled by changing the systematic vacuum degree according to the amount of the excreta. In particular, the amount of the excreta is tested by liquid level sensor and mass sensor. According to the amount of the excreta, the relationship between the excreta amount and the sewerage force is studied to provide proper propulsive force. In the other aspect, to provide variable vacuum degrees for different sanitation requirements, the suction and discharge system is designed with pneumatic vacuum ejector. On the basis of the static flow-rate characteristics and the vacuum generation model, the pressure response in the ejector circuit is studied by using the static flow rate characteristics of the ejector and air status equation. The relationship is obtained between supplied compressed air and systematic vacuum degree. When the compressed air is supplied to the ejector continuously, the systematic vacuum degree increases until the vacuum degree reaches the extreme value. Therefore, the variable systematic vacuum degree is obtained by controlling the compressed air supply of the ejector. To verify the effect of energy conservation, experiments are carried out in the artificial excreta collection, and the variable vacuum-degree design saves more than 30% of the energy supply. The energy conservation is realized effectively in the new vacuum sanitation systems with good application prospect. The proposed technology provides technological

  13. Modified ITER In-Vessel Viewing System

    International Nuclear Information System (INIS)

    Ahola, H.; Heikkinen, V.; Keraenen, K.; Suomela, J.

    2001-01-01

    The original ITER In-Vessel Viewing System (IVVS) prototype (Proc. of the 20th SOFT, vol. 2 (1998) 1051), which demonstrates the feasibility of linear fibre arrays for ITER in-vessel viewing, has been modified. In order to reduce the viewing time and to improve the image quality the beam dispersing mirrors was replaced by a diffractive optics element (DOE), which enhanced the laser illumination considerably. The performance of the system was tested using various target surfaces: the results obtained clearly indicate its adequacy for in-vessel viewing. Mechanical damage on smooth metal surfaces (scratches etc.) can be easily distinguished and the viewing resolution at a distance of 2 m is better than 1 mm. The IVVS has been re-designed to be compatible with the new ITER-FEAT. A conceptual study which covers all the functions and subsystems required for viewing has been completed. These results will be used to further modify the prototype: items to be tested include horizontal probe operation and laser illumination with an optical fibre

  14. Preliminary consideration of CFETR ITER-like case diagnostic system.

    Science.gov (United States)

    Li, G S; Yang, Y; Wang, Y M; Ming, T F; Han, X; Liu, S C; Wang, E H; Liu, Y K; Yang, W J; Li, G Q; Hu, Q S; Gao, X

    2016-11-01

    Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

  15. Preliminary consideration of CFETR ITER-like case diagnostic system

    International Nuclear Information System (INIS)

    Li, G. S.; Liu, Y. K.; Gao, X.; Yang, Y.; Wang, Y. M.; Ming, T. F.; Han, X.; Liu, S. C.; Wang, E. H.; Yang, W. J.; Li, G. Q.; Hu, Q. S.

    2016-01-01

    Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

  16. Nuclear aspects of diagnostics in RTO/RC ITER

    International Nuclear Information System (INIS)

    Walker, C.I.; Yamamoto, S.; Costley, A.; Kock, L. de; Ebisawa, K.; Janeschitz, G.; Khripunov, V.; Martin, E.; Vayakis, G.

    2000-01-01

    ITER (international thermonuclear experimental reactor) will be the first fusion device where the design of the plasma diagnostic systems will make extensive use of the materials and techniques developed in the nuclear technology field. The designs have to satisfy stringent requirements for tritium confinement, nuclear shielding and vacuum integrity. This paper introduces the requirements for diagnostics in the ITER long pulse, burning plasma environment, and addresses the impact of the reactor environment on the diagnostics and ancillary equipment. These systems necessarily require access to the plasma or first wall, which generally conflicts with the requirements of the basic machine. Holes are required through the first wall, primary shielding, containment boundaries and biological shielding. Components have a limited life and require maintenance. This paper describes the effect of the radiation environment on diagnostic design at different locations. Ex-vessel and in-vessel remote handling, hot cell refurbishment and tritium confinement are also described

  17. ITER nuclear components, preparing for the construction and R and D results

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.de; Akiba, M.; Barabaschi, P.; Barabash, V.; Chiocchio, S.; Daenner, W.; Elio, F.; Enoeda, M.; Ezato, K.; Federici, G.; Gervash, A.; Grebennikov, D.; Jones, L.; Kajiura, S.; Krylov, V.; Kuroda, T.; Lorenzetto, P.; Maruyama, S.; Merola, M.; Miki, N.; Morimoto, M.; Nakahira, M.; Ohmori, J.; Onozuka, M.; Rozov, V.; Sato, K.; Strebkov, Yu.; Suzuki, S.; Tanchuk, V.; Tivey, R.; Utin, Yu

    2004-08-01

    Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R and D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20 deg. or 30 deg. , on flow distribution tests of a two-channel model, on fabrication and testing of FW mock-ups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

  18. Vacuum system design considerations of the Los Alamos Accelerator Test Stand (ATS)

    International Nuclear Information System (INIS)

    Wilson, N.G.

    1986-01-01

    The accelerator test stand (ATS), in operation at the Los Alamos National Laboratory, includes a hydrogen ion source, low- and high-energy beam-transport sections, and a 425-MHz radio-frequency quadrupole (RFQ) linear accelerator. A 425-MHz drift-tube linac (DTL) and a powered ''buncher'' matching section have been constructed and will be installed on the ATS. The vacuum systems required for the various sections of the ATS are designed to provide: (1) high gas-load capability, as required in the ion source, and (2) high-vacuum capability in the high-power, radio-frequency accelerator sections (where fast vacuum-system response time is of importance) through the use of distributed, differential pumping as a principal vacuum-system feature. This paper describes properties of accelerator materials, vacuum-systems engineering and analysis, vacuum equipment used, and ATS vacuum-system performance

  19. ITER cooling systems

    International Nuclear Information System (INIS)

    Natalizio, A.; Hollies, R.E.; Sochaski, R.O.; Stubley, P.H.

    1992-06-01

    The ITER reference system uses low-temperature water for heat removal and high-temperature helium for bake-out. As these systems share common equipment, bake-out cannot be performed until the cooling system is drained and dried, and the reactor cannot be started until the helium has been purged from the cooling system. This study examines the feasibility of using a single high-temperature fluid to perform both heat removal and bake-out. The high temperature required for bake-out would also be in the range for power production. The study examines cost, operational benefits, and impact on reactor safety of two options: a high-pressure water system, and a low-pressure organic system. It was concluded that the cost savings and operational benefits are significant; there are no significant adverse safety impacts from operating either the water system or the organic system; and the capital costs of both systems are comparable

  20. Failure mode analysis of preliminary design of ITER divertor impurity monitor

    International Nuclear Information System (INIS)

    Kitazawa, Sin-iti; Ogawa, Hiroaki

    2016-01-01

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • Failure mode of DIM was prepared for RAMI analysis. • RAMI analysis on DIM was performed to reduce technical risks. - Abstract: The objective of the divertor impurity influx monitor (DIM) for ITER is to measure the parameters of impurities and hydrogen isotopes (tritium, deuterium, and hydrogen) in divertor plasma using visible and UV spectroscopic techniques in the 200–1000 nm wavelength range. In ITER, special provisions are required to ensure accuracy and full functionality of the diagnostic components under harsh conditions (high temperature, high magnetic field, high vacuum condition, and high radiation field). Japan Domestic Agency is preparing the preliminary design of the ITER DIM system, which will be installed in the upper, equatorial and lower ports. The optical and mechanical designs of the DIM are conducted to fit ITER’s requirements. The optical and mechanical designs meet the requirements of spatial resolution. Some auxiliary systems were examined via prototyping. The preliminary design of the ITER DIM system was evaluated by RAMI analysis. The availability of the designed system is adequately high to satisfy the project requirements. However, some equipment does not have certain designs, and this may cause potential technical risks. The preliminary design should be modified to reduce technical risks and to prepare the final design.

  1. Design and technical status of the EU contribution to ITER

    International Nuclear Information System (INIS)

    Gasparotto, Maurizio; Federici, Gianfranco; Casci, Federico Riccardo

    2009-01-01

    Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R and D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost. This paper provides an update of the design and technical status of EU contributions to ITER.

  2. Vacuum system and cleaning techniques in the FTU machines

    International Nuclear Information System (INIS)

    Alessandrini, C.; Apicella, M.L.; Ferro, C.

    1988-01-01

    FTU (Frascati Tokamak Upgrade) is a high magnetic field (8T) tokamak under construction at the Frascati Energy Research Center (ENEA). Its vacuum systems has been already manifactured and is presently being assembled. It consist of an all metallic fully welded vessel, pumped by six turbomolecular pumps. The vacuum system has been dimensioned to allow a base pressure lower than 2.6 x 10 -6 Pa. The paper reports the design philosophy of the vacuum system. The results of the cleaning techniques performed on a 1:1 scale toroidal sector of FTU are also presented and discussed

  3. Proceedings of the Japan-U.S. workshop P-118 on vacuum technologies for fusion devices

    International Nuclear Information System (INIS)

    Miyahara, A.

    1989-01-01

    Fusion community does not appreciate vacuum technologies to the same extent as accelerator community does. This is because, in the case of accelerators, in particular storage ring systems, the requirement of attaining ultrahigh vacuum in order to avoid collisional loss is well defined, on the other hand, it is not possible to define the requirement so precisely in the case of fusion devices. One of the reasons is that core plasma interacts with vessel wall so strongly and unpredictably that it becomes difficult to identify the role played by individual components. However, in the next step and the next generation machines like CIT, LHS, ITER, FER and NET, vacuum technologies would play more significant roles, because the CIT will introduce tritium in a vacuum vessel, and the aim of the ITER project is to demonstrate particle balance, namely, to achieve steady state operation with D-T fuel. The Japan-U.S. workshop P-118 was held at the Institute of Plasma Physics, Nagoya University, from August 1 to 5, 1988. 33 participants including 4 from the U.S. took part in the workshop. In the plenary session, 12 lectures were given, and also the topics-oriented session on pumping, gauging, remote maintenance, first wall, pump limiter, divertor and others was held. (K.I.)

  4. Absorbed decay-photon dose analysis of the IVVS/GDC plug in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Leichtle, D.; Serikov, A.; Fischer, U. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (DE). Inst. for Neutron Physics and Reactor Technology (INR)

    2011-07-01

    The In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit share a common port at the equatorial level of the ITER tokamak. The plug consists mainly of the IWS probe, capable of performing the laser-based in-vessel viewing and metrology, the GDC electrode, capable of producing glow discharge in the vacuum vessel during intermediate maintenance and wall conditioning periods, and their respective deployment systems to move the electrodes. The plug extends over a length of about 11 m from the GDC tip to the rear end at the bioshield level. At the present stage of the conceptual design a neutronics analysis has been requested to provide valuable input to the design strategy. To this end, a first assessment has been performed focusing on operational loads on the GDC electrode head in the so-called shielding position and on absorbed decay-photon dose rate levels in the structural components of the entire system. In this contribution we are reporting on the absorbed dose rates after the ITER life time irradiation at several cooling times. Gamma sources from activated materials of the IVVS/GDC and surrounding structures, like blanket, vacuum vessel, toroidal and poloidal field coils, have been taken into account. (orig.)

  5. The ITER poloidal field system

    Energy Technology Data Exchange (ETDEWEB)

    Wesley, J [General Atomics, San Diego, CA (USA); Beljakov, V; Kavin, A; Korshakov, V; Kostenko, A; Roshal, A; Zakharov, L [Kurchatov Inst. of Atomic Energy, Moscow (USSR); Bulmer, R; Kaiser, T; Miller, J R; Pearlstein, L D [Lawrence Livermore National Lab., CA (USA); Hogan, J [Oak Ridge National Lab., TN (USA); Kurihara, K; Shimomura, Y; Sugihara, M; Yoshino, R [Japan Atomic Energy Resea

    1990-12-15

    The ITER poloidal field (PF) system uses superconducting coils to provide the plasma equilibrium fields, slow equilibrium control and plasma flux linkage (V-s) needed for the ITER Operations and Research Program. Double-null (DN) divertor plasmas and operation scenarios for 22 MA Physics (high-Q/ignition) and 15 MA Technology (high-fluence testing) phases are provided. For 22 MA plasmas, total PF flux swing is 333 V-s. This provides inductive current drive (CD) for start-up with 66 V-s of resistive loss and 440-s (330-s minimum) sustained burn. The PF system also allows plasma start-up and shutdown scenarios, and can maintain the plasma configuration during burn over a range of current and pressure profiles. Other capabilities include increased plasma current (25 MA with inductive CD; 28 MA with non-inductive CD assist), divertor separatrix sweeping, and semi-DN and single-null plasmas.

  6. Experience with the TRIUMF Main Tank Vacuum Control System

    International Nuclear Information System (INIS)

    Sarkar, S.; Yandon, J.C.; Sievers, W.; Bennett, P.; Gurd, D.P.; Harmer, P.; Nelson, J.

    1993-01-01

    The TRIUMF Main Tank Vacuum Control System was upgraded in 1984. The earlier system, which consisted of a collection of hardwired relay logic boxes housed in three standard instrumentation racks, was replaced with a compact and flexible microprocessor-based control system. The user interface, previously distributed over the three racks, was consolidated into a single hardwired control and mimic panel. Since 1984, the Main Tank Vacuum System has undergone a series of changes in configuration and vacuum pumping hardware with necessary changes being implemented in the control system logic. Corresponding changes to the user interface were sometimes difficult to implement and in time exhausted the spare input/output capacity which had been built into the panel. The availability of inexpensive personal computers with adequate graphics capability and the ease of modifying, or adding to a programmable user interface precipitated the retirement of the hardwired panel and its replacement by a PC-based graphics user interface. System configuration, safety considerations, the hardware and the software implementation using the open-quote C close-quote programming language are described. The evolution of the control system and its performance, both over the years and in adapting to the vacuum system changes, are discussed

  7. Overview of magnetic control in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Zabeo, L., E-mail: luca.zabeo@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Ambrosino, G., E-mail: ambrosin@unina.it [CREATE/Universitá di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’informazione, Naples (Italy); Cavinato, M., E-mail: mario.cavinato@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla 2, Torres Diagonal Litoral - B3, 08019 Barcelona (Spain); Gribov, Y., E-mail: yuri.gribov@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Kavin, A., E-mail: kavina@sintez.niiefa.spb.su [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Lukash, V., E-mail: lukash@nfi.kiae.ru [Kurchatov Institute, Moscow (Russian Federation); Mattei, M., E-mail: massimiliano.mattei@unina2.it [CREATE/Seconda Universitá di Napoli, Dip. Ingegneria Industriale e dell’informazione, Naples (Italy); Pironti, A., E-mail: pironti@unina.it [CREATE/Seconda Universitá di Napoli, Dip. Ingegneria Industriale e dell’informazione, Naples (Italy); Snipes, J.A., E-mail: joseph.snipes@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Vayakis, G., E-mail: george.vayakis@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Winter, A., E-mail: axel.winter@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France)

    2014-05-15

    ITER is targeting Q = 10 with 500 MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination. Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios. Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design. Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the

  8. LHC : The World's Largest Vacuum Systems being commissioned at CERN

    CERN Document Server

    Jiménez, J M

    2008-01-01

    When it switches on in 2008, the 26.7 km Large Hadron Collider (LHC) at CERN, will have the world's largest vacuum system operating over a wide range of pressures and employing an impressive array of vacuum technologies. This system is composed by 54 km of UHV vacuum for the circulating beams and 50 km of insulation vacuum around the cryogenic magnets and the liquid helium transfer lines. Over the 54 km of UHV beam vacuum, 48 km of this are at cryogenic temperature (1.9 K). The remaining 6 km of beam vacuum containing the insertions for "cleaning" the proton beams, radiofrequency cavities for accelerating the protons as well as beam-monitoring equipment is at ambient temperature and uses non-evaporable getter (NEG) coatings - a vacuum technology that was born and industrialized at CERN. The pumping scheme is completed using 780 ion pumps to remove noble gases and to provide pressure interlocks to the 303 vacuum safety valves. Pressure readings are provided by 170 Bayard-Alpert gauges and 1084 gauges (Pirani a...

  9. Iterative solution of large linear systems

    CERN Document Server

    Young, David Matheson

    1971-01-01

    This self-contained treatment offers a systematic development of the theory of iterative methods. Its focal point resides in an analysis of the convergence properties of the successive overrelaxation (SOR) method, as applied to a linear system with a consistently ordered matrix. The text explores the convergence properties of the SOR method and related techniques in terms of the spectral radii of the associated matrices as well as in terms of certain matrix norms. Contents include a review of matrix theory and general properties of iterative methods; SOR method and stationary modified SOR meth

  10. The vacuum system of the Karlsruhe magnetic spectrograph 'Little John'

    International Nuclear Information System (INIS)

    Buschmann, J.; Gils, H.J.; Jelitto, H.; Krisch, J.; Ludwig, G.; Manger, D.; Rebel, H.; Seith, W.; Zagromski, S.

    1985-02-01

    The vacuum equipment of the magnetic spectrograph Little John is described. The system is characterized by the following special features: The sliding exit flange of the target chamber can be moved to the desired angle of observation without affecting the high vacuum. The pressure maintained is less by a factor of ten than the pressure in the incoming beam tubing. The vacuum system is divided into several separate pumping sections. Ground loops are strictly avoided. All actual states of relevance are fed back to the control panels. The vacuum installation is protected by hardware interlocking systems as well as by a real time program written in FORTRAN in cooperation with CAMAC interfacing. (orig.) [de

  11. ITER neutral beam system

    International Nuclear Information System (INIS)

    Mondino, P.L.; Di Pietro, E.; Bayetti, P.

    1999-01-01

    The Neutral Beam (NB) system for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration with the tokamak and with the rest of the plant. Operational requirements and maintainability have been considered in the design. The paper considers the integration with the tokamak, discusses design improvements which appear necessary and finally notes R and D progress in key areas. (author)

  12. Integration of the ITER diagnostic plant systems with CODAC

    International Nuclear Information System (INIS)

    Simrock, S.; Barnsley, R.; Bertalot, L.; Hansalia, C.; Klotz, W.D.; Makijarvi, P.; Reichle, R.; Vayakis, G.; Yonekawa, I.; Walker, C.; Wallander, A.; Walsh, M.; Winter, A.

    2011-01-01

    ITER requires extensive diagnostic systems in order to meet the requirements for machine operation, protection, plasma control and physics studies. The realization of these systems is a major challenge not only because of the harsh environment and the nuclear requirements but also with respect to Instrumentation and Control (I and C) of all the 59 diagnostics plants. The Plant Systems I and C are mostly 'in-kind', i.e. procured by the seven ITER Domestic Agencies (DAs), while the Central I and C Systems are 'in-fund', i.e. procured by ITER Organization (IO). Standardization of Plant Systems I and C is of primary importance and has been one of the highest priority tasks of CODAC. The standards are published in the Plant Control Design Handbook (PCDH) which will be followed to ensure a homogeneous design, guarantee high availability and simplify maintenance and support future upgrades. Most important for a successful commissioning and operation of the ITER facility are the concepts of interfacing the diagnostics plant systems with CODAC and the standards for instrumentation and control which must be followed all contributing parties. In this paper, we will elaborate on the concepts of interfacing the diagnostics plant systems with CODAC and the standards that must be followed for the design.

  13. 242-A evaporator vacuum condenser system

    International Nuclear Information System (INIS)

    Smith, V.A.

    1994-01-01

    This document is written for the 242-A evaporator vacuum condenser system (VCS), describing its purpose and operation within the evaporator. The document establishes the operating parameters specifying pressure, temperature, flow rates, interlock safety features and interfacing sub-systems to support its operation

  14. Seismic analysis of ITER multi-purpose deployer

    International Nuclear Information System (INIS)

    Manuelraj, Manoah Stephen; Gotewal, Krishan Kumar; Dutta, Pramit; Rastogi, Naveen; Choi, Chang-Hwan; Tesini, Alessandro

    2015-01-01

    The Multi-Purpose Deployer (MPD) is a general purpose ITER in-vessel remote handling (RH) system. The MPD will perform various in-vessel maintenance tasks such as dust and tritium inventory control, in-service inspection, leak localization and in-vessel diagnostics maintenance. The main handling equipment, called as the MPD Transporter, consists of a series of linked bodies, which provide anchoring to the vacuum vessel port and an articulated multi-degree of freedom motion to perform the aforementioned tasks. The target payload for the MPD Transporter is 2 tons. The total length is 16.6 m and 18.1 m for short and long configuration respectively, while the total weight of the system is about 25.5 tons including the payload. During the in-vessel operations, the structural integrity of the system should be guaranteed against various operational and seismic loads. This paper presents the seismic structural analysis results of the concept design of the MPD Transporter. Static structural, modal and frequency response spectrum analyses have been performed to verify the structural integrity of the MPD itself, and to provide reaction loads to the interfacing systems such as vacuum vessel and cask. The analyses are carried out by using the ANSYS. The first analysis iteration was carried out for the reference design of the MPD Transporter, which showed stresses higher than the permissible limit. Structural optimizations and reinforcements were performed for individual bodies referring the stress levels in each body, and a reinforced design was proposed. The reinforced design satisfies the required structural criteria in terms of general global stresses. Though local stress concentrations were observed, it can be solved in the next design phase by further local reinforcements or proper material choice. (author)

  15. European technology activities to prepare for ITER component procurement

    International Nuclear Information System (INIS)

    Gasparotto, M.

    2006-01-01

    Over the past few years the technology activities of the European fusion programme have principally been devoted to: a) the completion of design and R (and) D studies in preparation for the procurement of ITER systems and components in close collaboration with the ITER team and according to the ITER design and schedule; b) provision of support to European industry and associations in key areas of fusion R (and) D to ensure a competitive and timely approach to the planned procurement. The EU contribution to ITER design and R (and) D activities has been maintained at a significant level with the objectives of: · continuing, and in some areas expanding, the effort in areas where design and R (and)D are still required: in particular in Machine Assembly, Remote Handling, ITER Test Blanket Modules, Diagnostics, Heating and Current Drive Systems. · continuing and completing manufacturing R (and)D to determine the most technically and cost affective manufacturing methods for ITER components to be built in Europe. · preparing new test facilities needed during ITER construction (DIPOLE, HELOKA, DTP-2, ECRH Test Facility, Fatigue Testing Facility). · supporting the European site preparation process and the preparation of safety and licensing documentation for ITER in Cadarache. · maintaining support to EU industries in R (and) D activities of relevance to fusion. To support the ITER Design activities and to prepare for the provision of timely answers to key issues, which may be raised during the ITER design review, support from specialized companies has been set-up in the fields of Civil and General Plant Engineering, Mechanical Engineering / Components, Mechanical Engineering / Systems (and) Plants, Remote Handling (and) Assembly, Electrical Engineering, Nuclear Safety Engineering. In recent years major efforts have been directed towards the technology development of the ITER components for which procurement can be launched during the first years of the construction

  16. 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)

  17. A note on iterated function systems with discontinuous probabilities

    International Nuclear Information System (INIS)

    Jaroszewska, Joanna

    2013-01-01

    Highlights: ► Certain iterated function system with discontinuous probabilities is discussed. ► Existence of an invariant measure via the Schauder–Tychonov theorem is established. ► Asymptotic stability of the system under examination is proved. -- Abstract: We consider an example of an iterated function system with discontinuous probabilities. We prove that it posses an invariant probability measure. We also prove that it is asymptotically stable provided probabilities are positive

  18. A Conceptual Design and Structural Analysis for ITER Mid-plane Brace Tools

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Kihak; Robert, Shaw [ITER Organization, St Paul lez Durance Cedex (France)

    2010-10-15

    The ITER, International Thermonuclear Experimental Reactor, Tokamak machine is mainly composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 .deg. sectors. Each VV/TFCs/VVTS 40 .deg. sector is made up of one 40 .deg. VV, two 20 .deg. TFCs and associated VVTS segments. The ITER Tokamak assembly tools are purpose-built tools to assemble the ITER Tokamak machine which includes the cryostat and the components contained therein. Based on the design description document prepared by the IO (ITER international organization), Korea has carried out the conceptual design of assembly tools with IO cooperation. The 40 .deg. sector assemblies attached mid-plane brace tools sub-assembled at assembly hall are transferred to Tokamak hall using the lifting tool operated by Tokamak main cranes. The sector sub-assembly tools are composed of the upending tool, the sector sub-assembly tool, the sector lifting tool and the vacuum vessel support and mid-plane brace tools. The mid-plane brace tool is assembled to inner surface of VV and TFCs in phase of sector sub-assembly after completion of all sector components. VV, TFC and VVTS are separated fully before completion of 9 sectors at Tokamak in-pit. In this paper the mid-plane brace tools is introduced about function, structure and status of research and development are also described

  19. Structural analysis of the ITER Divertor toroidal rails

    Energy Technology Data Exchange (ETDEWEB)

    Viganò, F., E-mail: Fabio.Vigano@LTCalcoli.it [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Escourbiac, F.; Gicquel, S.; Komarov, V. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Lucca, F. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ngnitewe, R. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy)

    2013-10-15

    The Divertor is one of the most technically challenging components of the ITER machine, which has the main function of extracting the power conducted in the scrape-off layer while maintaining the plasma purity. There are 54 Divertor cassettes installed in the vacuum vessel (VV). Each cassette body (CB) is fastened on the inner and outer concentric Divertor toroidal rails. The comprehensive assessment (in accordance with the Structural Design Criteria for ITER In-vessel Components: ITER SDC-IC) of the Divertor toroidal rails has been performed during design activity based on performing of thermal and stress analyses at operating conditions of neutron stage of ITER operation. This paper outlines the engineering aspects of the ITER Divertor toroidal rails and focuses on some critical regions of the present design highlighted by the performed structural assessment. The structural assessment has been performed with help of using Finite Element (FE) Abaqus code and based on criteria given by ITER SDC-IC.

  20. Erosion evaluation capability of the IVVS for ITER applications

    Energy Technology Data Exchange (ETDEWEB)

    Pollastrone, Fabio, E-mail: fabio.pollastrone@enea.it [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Ferri de Collibus, Mario; Florean, Marco; Francucci, Massimo; Mugnaini, Giampiero; Neri, Carlo; Rossi, Paolo [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

    Highlights: •High resolution laser radar range images for hostile environment (IVVS). •Evaluation of the erosion on the surface scanned by IVVS laser radar. •Erosion evaluation procedure and software. •Test and results of the erosion evaluation procedure. -- Abstract: In ITER it is foreseen the use of the In Vessel Viewing System (IVVS), whose scanning head is a 3D laser imaging system able to obtain high-resolution intensity and range images in hostile environments. The IVVS will be permanently installed into a port extension, therefore it has to be compliant with ITER primary vacuum requirements. In the frame of a Fusion for Energy Grant, an investigation of the expected IVVS metrology performances was required to evaluate the device capability to detect erosions on ITER first wall and divertor and to estimate the amount of eroded material. In ENEA Frascati laboratories, an IVVS probe prototype was developed along with a method and a computational procedure applied to a reference erosion plate target simulating ITER vessel components and their possible erosions. Experimental tests were carried out by this system performing several scans of the reference target with different incidence angles, estimating the eroded volume and comparing this volume with its true value. A dedicated study has been also done by changing the power of the laser source; a discussion about the quality of the 3D laser images is reported. The main results obtained during laboratory tests and data processing are presented and discussed.

  1. Vacuum pumping system for the JT-60 radio-frequency heating system

    International Nuclear Information System (INIS)

    Yokokura, Kenji; Ikeda, Yoshitaka; Imai, Tuyoshi; Suganuma, Kazuaki; Nagashima, Takashi

    1988-01-01

    The basic design requirements set up for the JT-60 radio-frequency heating system included: (1) rapid pumping of gas released upon application of a radio-frequency power to maintain the pressure in the launchers at 10 -2 - 10 -3 Pa or less, (2) incorporation of a gas analysis system that can operate under a strong field and high pressure (>10 -2 Pa) to permit remote controlled data collection and processing, and (3) low cost, multiple functions and high reliability. The vacuum pumping system, consisting of three units for low hybrid radio-frequency (LHRF) and one unit for ion cyclotron radio-frequency (ICRF), is connected to each launcher provided at the four ports of JT-60. The LHRF unit is composed of a main pump, an alumina joint for electrical insulation from the launcher, a metallic gate valve for isolation from the JT-60 vacuum region, and various vacuum gauges. Only a turbo-molecular pump is used for the ICRF system because a large-scale differential pumping is not required. A gas measuring system is incorporated which consists of a mass filter, personal computer, turbo-molecular pump, and variable flow valve equipped with an APG control. This system is designed to identify and make use of gas impurities released during the launcher aging process. The control system employed consists of a personal computer, interlock control board, data logger and other devices such as vacuum gages. (Nogami, K.)

  2. Design of ITER neutron monitor using micro fission chambers

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi; Johnson, L.C.; Walker, C.

    1998-08-01

    We are designing micro fission chambers, which are pencil size gas counters with fissile material inside, to be installed in the vacuum vessel as neutron flux monitors for ITER. We found that the 238 U micro fission chambers are not suitable because the detection efficiency will increase up to 50% in the ITER life time by breading 239 Pu. We propose to install 235 U micro fission chambers on the front side of the back plate in the gap between adjacent blanket modules and behind the blankets at 10 poloidal locations. One chamber will be installed in the divertor cassette just under the dome. Employing both pulse counting mode and Campbelling mode in the electronics, we can accomplish the ITER requirement of 10 7 dynamic range with 1 ms temporal resolution, and eliminate the effect of gamma-rays. We demonstrate by neutron Monte Carlo calculation with three-dimensional modeling that we avoid those detection efficiency changes by installing micro fission chambers at several poloidal locations inside the vacuum vessel. (author)

  3. 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

  4. Venturi vacuum systems for hypobaric chamber operations.

    Science.gov (United States)

    Robinson, R; Swaby, G; Sutton, T; Fife, C; Powell, M; Butler, B D

    1997-11-01

    Physiological studies of the effects of high altitude on man often require the use of a hypobaric chamber to simulate the reduced ambient pressures. Typical "altitude" chambers in use today require complex mechanical vacuum systems to evacuate the chamber air, either directly or via reservoir system. Use of these pumps adds to the cost of both chamber procurement and maintenance, and service of these pumps requires trained support personnel and regular upkeep. In this report we describe use of venturi vacuum pumps to perform the function of mechanical vacuum pumps for human and experimental altitude chamber operations. Advantages of the venturi pumps include their relatively low procurement cost, small size and light weight, ease of installation and plumbing, lack of moving parts, and independence from electrical power sources, fossil fuels and lubricants. Conversion of three hyperbaric chambers to combined hyper/hypobaric use is described.

  5. Progress in Development of the ITER Plasma Control System Simulation Platform

    Science.gov (United States)

    Walker, Michael; Humphreys, David; Sammuli, Brian; Ambrosino, Giuseppe; de Tommasi, Gianmaria; Mattei, Massimiliano; Raupp, Gerhard; Treutterer, Wolfgang; Winter, Axel

    2017-10-01

    We report on progress made and expected uses of the Plasma Control System Simulation Platform (PCSSP), the primary test environment for development of the ITER Plasma Control System (PCS). PCSSP will be used for verification and validation of the ITER PCS Final Design for First Plasma, to be completed in 2020. We discuss the objectives of PCSSP, its overall structure, selected features, application to existing devices, and expected evolution over the lifetime of the ITER PCS. We describe an archiving solution for simulation results, methods for incorporating physics models of the plasma and physical plant (tokamak, actuator, and diagnostic systems) into PCSSP, and defining characteristics of models suitable for a plasma control development environment such as PCSSP. Applications of PCSSP simulation models including resistive plasma equilibrium evolution are demonstrated. PCSSP development supported by ITER Organization under ITER/CTS/6000000037. Resistive evolution code developed under General Atomics' Internal funding. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

  6. Baseline Architecture of ITER Control System

    Science.gov (United States)

    Wallander, A.; Di Maio, F.; Journeaux, J.-Y.; Klotz, W.-D.; Makijarvi, P.; Yonekawa, I.

    2011-08-01

    The control system of ITER consists of thousands of computers processing hundreds of thousands of signals. The control system, being the primary tool for operating the machine, shall integrate, control and coordinate all these computers and signals and allow a limited number of staff to operate the machine from a central location with minimum human intervention. The primary functions of the ITER control system are plant control, supervision and coordination, both during experimental pulses and 24/7 continuous operation. The former can be split in three phases; preparation of the experiment by defining all parameters; executing the experiment including distributed feed-back control and finally collecting, archiving, analyzing and presenting all data produced by the experiment. We define the control system as a set of hardware and software components with well defined characteristics. The architecture addresses the organization of these components and their relationship to each other. We distinguish between physical and functional architecture, where the former defines the physical connections and the latter the data flow between components. In this paper, we identify the ITER control system based on the plant breakdown structure. Then, the control system is partitioned into a workable set of bounded subsystems. This partition considers at the same time the completeness and the integration of the subsystems. The components making up subsystems are identified and defined, a naming convention is introduced and the physical networks defined. Special attention is given to timing and real-time communication for distributed control. Finally we discuss baseline technologies for implementing the proposed architecture based on analysis, market surveys, prototyping and benchmarking carried out during the last year.

  7. Design of cryo-vacuum system for MW neutral beam injector

    International Nuclear Information System (INIS)

    Hu Chundong; Xie Yuanlai

    2010-01-01

    Neutral beam injector is an equipment that is used to produce and then to neutralize high energetic particle beam. A neutral beam injector (EAST-NBI) with MW magnitude neutral beam power is considered to be developed to support the EAST physical research. The requirements for vacuum system were analyzed after introducing the principle of EAST-NBI. A differential vacuum system structure was chosen after analyzing the performance of different vacuum pumping system structure. The gas sources and their characteristics were analyzed, and two inserted type cryocondensation pumps were chosen as main vacuum pump. The schematic structure of the two cryocondensation pump with pumping area 8 m 2 and 6 m 2 were given and their cooling method and temperature control mode were determined. (authors)

  8. European Technological Effort in Preparation of ITER Construction

    International Nuclear Information System (INIS)

    Andreani, Roberto

    2005-01-01

    Europe has started since the '80s with the preparatory work done on NET, the Next European Torus, the successor of JET, to prepare for the construction of the next generation experiment on the road to the fusion reactor. In 2000 the European Fusion Development Agreement (EFDA) has been signed by sixteen countries, including Switzerland, not a member of the Union. Now the signatory countries have increased to twenty-five. A vigorous programme of design and R and D in support of ITER construction has been conducted by EFDA through the coordinated effort of the national institutes and laboratories supported financially, in the framework of the VI European Framework Research Programme (2002-2006), by contracts of association with EURATOM. In the last three years, with the expenditure of 160 M[Euro], the accent has been particularly put on the preparation of the industrial manufacturing activities of components and systems for ITER. Prototypes and manufacturing methods have been developed in all the main critical areas of machine construction with the objective of providing sound and effective solutions: vacuum vessel, toroidal field coils, poloidal field coils, remote handling equipment, plasma facing components and divertor components, electrical power supplies, generators and power supplies for the Heating and Current Drive Systems and other minor subsystems.Europe feels to be ready to host the ITER site and to provide adequate support and guidance for the success of construction to our partners in the ITER collaboration, wherever needed

  9. Manufacturing progress on the first sector and lower ports for ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, H.J., E-mail: hjahn@nfri.re.kr [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kim, H.S.; Kim, G.H.; Park, C.K.; Hong, G.H.; Jin, S.W.; Lee, H.G.; Jung, K.J. [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Lee, J.S.; Kim, T.S.; Won, J.G.; Roh, B.R.; Park, K.H. [Hyundai Heavy Industries Co. Ltd., Ulsan 682-792 (Korea, Republic of); Sa, J.W.; Choi, C.H.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France)

    2016-11-01

    Highlights: • All manufacturing drawings of the first sector of VV have been completed. • Full scale mock-ups have been constructed to verify fabrication procedure. • Qualifications for welding and forming are done and for NDE are ongoing. • Manufacturing progress is around 40% for the sector and LPSE up to the end of 2015. - Abstract: Manufacturing design of Korean sectors and ports for the ITER Vacuum Vessel (VV) has been developed to comply with the tight tolerance and severe inspection requirements. The first VV sector and lower ports are being fabricated slowly under strict regulations after verification using several real scale mock-ups and qualifications for welding, forming and NDE. During three years after start of fabrication, manufacturing progress on four poloidal segments of the first sector is that (1) all inner shells were welded, (2) forgings for complicate components have been machined, (3) port stubs and poloidal T-ribs were assembled, and (4) machined components are welded on the inner shells by narrow-gap TIG welding and electron beam welding. The progress of lower ports is that (1) inner shells of stub extensions were bent and treated with heat, (2) T-ribs were fabricated and examined by qualified phased array UT, (3) supporting pads and gussets have been machined, and (4) inner shells are assembled with T-ribs and machined forgings. The progress rate of manufacturing is around 40% up to the end of 2015 for the first sector and lower port stub extensions.

  10. Architectural concept for the ITER Plasma Control System

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Humphreys, D., E-mail: humphreys@fusion.gat.com [General Atomics, San Diego, CA (United States); Raupp, G., E-mail: Gerhard.Raupp@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Schuster, E., E-mail: schuster@lehigh.edu [Lehigh University, Bethlehem, PA (United States); Snipes, J., E-mail: Joseph.Snipes@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France); De Tommasi, G., E-mail: detommas@unina.it [CREATE/Università di Napoli Federico II, Napoli (Italy); Walker, M., E-mail: walker@fusion.gat.com [General Atomics, San Diego, CA (United States); Winter, A., E-mail: Axel.Winter@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France)

    2014-05-15

    The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

  11. Architectural concept for the ITER Plasma Control System

    International Nuclear Information System (INIS)

    Treutterer, W.; Humphreys, D.; Raupp, G.; Schuster, E.; Snipes, J.; De Tommasi, G.; Walker, M.; Winter, A.

    2014-01-01

    The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

  12. Simulation-based design process for the verification of ITER remote handling systems

    International Nuclear Information System (INIS)

    Sibois, Romain; Määttä, Timo; Siuko, Mikko; Mattila, Jouni

    2014-01-01

    Highlights: •Verification and validation process for ITER remote handling system. •Simulation-based design process for early verification of ITER RH systems. •Design process centralized around simulation lifecycle management system. •Verification and validation roadmap for digital modelling phase. -- Abstract: The work behind this paper takes place in the EFDA's European Goal Oriented Training programme on Remote Handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. One of the projects of this programme focuses on the verification and validation (V and V) of ITER RH system requirements using digital mock-ups (DMU). The purpose of this project is to study and develop efficient approach of using DMUs in the V and V process of ITER RH system design utilizing a System Engineering (SE) framework. Complex engineering systems such as ITER facilities lead to substantial rise of cost while manufacturing the full-scale prototype. In the V and V process for ITER RH equipment, physical tests are a requirement to ensure the compliance of the system according to the required operation. Therefore it is essential to virtually verify the developed system before starting the prototype manufacturing phase. This paper gives an overview of the current trends in using digital mock-up within product design processes. It suggests a simulation-based process design centralized around a simulation lifecycle management system. The purpose of this paper is to describe possible improvements in the formalization of the ITER RH design process and V and V processes, in order to increase their cost efficiency and reliability

  13. Construction of vacuum system for Tristan accumulation ring

    International Nuclear Information System (INIS)

    Ishimaru, H.; Horikoshi, G.; Kobayashi, M.; Kubo, T.; Mizuno, H.; Momose, T.; Narushima, K.; Watanabe, H.; Yamaguchi, H.

    1983-01-01

    An all aluminum-alloy vacuum system for the TRISTAN accumulation ring is now under construction. Aluminum and aluminum alloys are preferred materials for ultrahigh vacuum systems of large electron storage rings because of their good thermal conductivity, extremely low outgassing rate, and low residual radioactivity. Vacuum beam chambers for the dipole and quadrupole magnets are extruded using porthole dies. The aluminum alloy 6063-T6 provides superior performance in extrusion. For ultrahigh vacuum performance, a special extrusion technique is applied which, along with the outgassing procedure used, is described in detail. Aluminum alloy 3004 seamless elliptical bellows are inserted between the dipole and quadrupole magnet chambers. These bellows are produced by the hydraulic forming of a seamless tube. The seamless bellows and the beam chambers are joined by fully automatic welding. The ceramic chambers for the kicker magnets, the fast bump magnets, and the slow beam intensity monitor are inserted in the aluminum alloy beam chambers. The ceramic chamber (98% alumina) and elliptical bellows are brazed with brazing sheets (4003-3003-4003) in a vacuum furnace. The brazing technique is described. The inner surface of the ceramic chamber is coated with a TiMo alloy by vacuum evaporation to permit a smooth flow of the RF wall current. Other suitable aluminum alloy components, including fittings, feedthroughs, gauges, optical windows, sputter ion pumps, turbomolecular pumps, and valves have been developed; their fabrication is described

  14. Evaluation of ISABELLE full cell ultra high vacuum system

    International Nuclear Information System (INIS)

    Foerster, C.L.; Briggs, J.; Chou, T.S.; Stattel, P.

    1980-01-01

    The ISABELLE Full Cell Vacuum System consisting of a 40 m long, by 8.8 cm diameter stainless steel tube pumped by seven pumping stations was assembled and processed for 10 -12 Torr operation. Evaluation and testing of the system and its sub-assemblies has been completed. Detail design of system components and the determination of the conditioning process was completed. The best procedure to rough pump, leak test, vacuum bake the system, condition pumps, degas gauges, turn on ion pumps and flash sublimation pumps was established. Pressures below 2 x 10 -11 Torr are now routinely achieved in normal operation of the Full Cell. This includes pump down after replacement of various components and pump down after back fill with moist unfiltered air. The techniques developed for the Full Cell will be used to build the ISABELLE Ultra High Vacuum System

  15. Proceedings of the international conference on vacuum science and technology and SRS vacuum systems. V.1: accelerators and SRS systems

    International Nuclear Information System (INIS)

    Venkatramani, N.; Sinha, A.K.

    1995-01-01

    An International Conference on Vacuum Science and Technology, INCOVAST-95 was held during January 30 - February 2, 1995 at the Centre for Advanced Technology (CAT), Indore under the aegis of the Indian Vacuum Society. Centre for Advanced Technology has a major programme of design and construction of a 450 MeV electron storage ring, synchrotron radiation source Indus-1 followed by the 1.25 GeV Indus-2. To match the activities at the centre, the present conference had ultrahigh vacuum for Synchrotron Radiation Sources (SRSs) as the main theme. Three major topics, namely accelerators and SRS systems, thin films and surfaces, vacuum components and applications were covered in detail. A short summary of the discussions is also included in the proceedings. Papers relevant to INIS are indexed separately

  16. Conceptual design of SC magnet system for ITER, (6)

    International Nuclear Information System (INIS)

    Yoshida, Kiyoshi; Sugimoto, Makoto; Tsuji, Hiroshi

    1991-08-01

    The International Thermonuclear Experimental Reactor (ITER) is an experimental thermonuclear tokamak reactor in order to test the basic physics performance and technologies. The conceptual design activity (CDA) of ITER required the joint work at a technical site at the Max Plank Institute for Plasma Physics in the Garching, Germany from 1988 to 1990. The technical proposals from Japan were summarized by the Fusion Experimental Reactor (FER) Team and the Superconducting Magnet Laboratory of the Japan Atomic Energy Research Institute (JAERI). This paper describes the Japanese contributions of the R and D proposals to the magnet system for the ITER. These proposals were discussed in ITER CDA design team and summarized in ITER Technical report No. 20. The development program of Toroidal Field Coil is basically proposed from Japan with the design and analysis reports. The Japanese proposals are almost adopted in the ITER Long-Term R and D program. (author)

  17. Vacuum Acceptance Tests for the UHV Room Temperature Vacuum System of the LHC during LS1

    CERN Document Server

    Cattenoz, G; Bregliozzi, G; Calegari, D; Gallagher, J; Marraffa, A; Chiggiato, P

    2014-01-01

    During the CERN Large Hadron Collider (LHC) first long shut down (LS1), a large number of vacuum tests are carried out on consolidated or newly fabricated devices. In such a way, the vacuum compatibility is assessed before installation in the UHV system of the LHC. According to the equipment’s nature, the vacuum acceptance tests consist in functional checks, leak test, outgassing rate measurements, evaluation of contaminants by Residual Gas Analysis (RGA), pumping speed measurements and qualification of the H2 sticking probability of Non-Evaporable-Getter (NEG) coating. In this paper, the methods used for the tests and the acceptance criteria are described. A summary of the measured vacuum characteristics for the tested components is also given.

  18. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

    Mardani, M.; Amrollahi, R.; Koohestani, S.

    2012-01-01

    Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

  19. Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

    International Nuclear Information System (INIS)

    Gargiulo, L.; Cordier, J.J.; Friconneau, J.P.; Grisolia, C.; Palmer, J.D.; Perrot, Y.; Samaille, F.

    2007-01-01

    The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel remote handling inspection system using a long reach, limited payload carrier. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development will allow close inspection of the Tore Supra plasma facing components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be simulated on Tore Supra through the deuterium inventory under long-time plasma discharges. The in situ leakage localisation of a damaged plasma facing component is also one of the major ITER maintenance challenges that could use remote handling inspection tools

  20. Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

    Energy Technology Data Exchange (ETDEWEB)

    Gargiulo, L. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)], E-mail: laurent.gargiulo@cea.fr; Cordier, J.J. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Friconneau, J.P. [CEA-LIST Robotics and Interactive Systems Unit, BP6 F-92265 Fontenay aux Roses Cedex (France); Grisolia, C. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Palmer, J.D. [EFDA CSU, Max-Planck-Institut fuer Plasma Physik Boltzmannstr. 2, D-85748 Garching (Germany); Perrot, Y. [CEA-LIST Robotics and Interactive Systems Unit, BP6 F-92265 Fontenay aux Roses Cedex (France); Samaille, F. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)

    2007-10-15

    The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel remote handling inspection system using a long reach, limited payload carrier. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development will allow close inspection of the Tore Supra plasma facing components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be simulated on Tore Supra through the deuterium inventory under long-time plasma discharges. The in situ leakage localisation of a damaged plasma facing component is also one of the major ITER maintenance challenges that could use remote handling inspection tools.

  1. 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

  2. Data archiving system implementation in ITER's CODAC Core System

    International Nuclear Information System (INIS)

    Castro, R.; Abadie, L.; Makushok, Y.; Ruiz, M.; Sanz, D.; Vega, J.; Faig, J.; Román-Pérez, G.; Simrock, S.; Makijarvi, P.

    2015-01-01

    Highlights: • Implementation of ITER's data archiving solution. • Integration of the solution into CODAC Core System. • Data archiving structure. • High efficient data transmission into fast plant system controllers. • Fast control and data acquisition in Linux. - Abstract: The aim of this work is to present the implementation of data archiving in ITER's CODAC Core System software. This first approach provides a client side API and server side software allowing the creation of a simplified version of ITERDB data archiving software, and implements all required elements to complete data archiving flow from data acquisition until its persistent storage technology. The client side includes all necessary components that run on devices that acquire or produce data, distributing and streaming to configure remote archiving servers. The server side comprises an archiving service that stores into HDF5 files all received data. The archiving solution aims at storing data coming for the data acquisition system, the conventional control and also processed/simulated data.

  3. PEP-II vacuum system pressure profile modeling using EXCEL

    International Nuclear Information System (INIS)

    Nordby, M.; Perkins, C.

    1994-06-01

    A generic, adaptable Microsoft EXCEL program to simulate molecular flow in beam line vacuum systems is introduced. Modeling using finite-element approximation of the governing differential equation is discussed, as well as error estimation and program capabilities. The ease of use and flexibility of the spreadsheet-based program is demonstrated. PEP-II vacuum system models are reviewed and compared with analytical models

  4. ITER fuel cycle systems layout

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-10-01

    The ITER fuel cycle building (FCB) will contain the following systems: fuel purification - permeator based; fuel purification - molecular sieves; impurity treatment; waste water storage and treatment; isotope separation; waste water tritium extraction; tritium extraction from solid breeder; tritium extraction from test modules; tritium storage, shipping and receiving; tritium laboratory; atmosphere detritiation systems; fuel cycle control centre; tritiated equipment maintenance space; control maintenance space; health physics laboratory; access, access control and facilities. The layout of the FCB and the requirements for these systems are described. (10 figs.)

  5. A proposal for the ITER remote participation system in Japan

    International Nuclear Information System (INIS)

    Nagayama, Y.; Emoto, M.; Kozaki, Y.; Nakanishi, H.; Sudo, S.; Yamamoto, T.; Hiraki, K.; Urushidani, S.

    2010-01-01

    This paper presents a proposal of the remote participation system for the international thermonuclear experimental reactor (ITER). The object of this paper is to clarify technical issues to analyze the ITER data safely and conveniently. The Japanese case is considered as an example, but technologies presented here can be used worldwide. Major technical issues are as follows: (1) the long distance data transfer; (2) the massive data server; (3) the secure network; (4) the convenient and fast data analysis system. Raw data of ITER can be transferred from France to Japan in a short time by optimizing TCP/IP parameters. The virtual private network (VPN) technology provides a secure environment of the data mirroring and the distributed computation. The analysis server with the WEB user interface enables physicists to analyze the ITER data from the Internet. Streaming data, such as plasma parameters in the steady state, video and sound of the ITER plasma and the status of experiment, which provides feeling of reality, are delivered by using the multi-cast technology. These technologies are being developed in SNET, which is a virtual laboratory for Japanese fusion community. International collaboration is required to develop a global distributed file system and a data analysis system further.

  6. A proposal for the ITER remote participation system in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Nagayama, Y., E-mail: nagayama.yoshio@nifs.ac.j [National Institute for Fusion Science, 322-6 Oroshi, Toki 509-5292 (Japan); Emoto, M.; Kozaki, Y.; Nakanishi, H.; Sudo, S.; Yamamoto, T. [National Institute for Fusion Science, 322-6 Oroshi, Toki 509-5292 (Japan); Hiraki, K. [Graduate School of Information Science and Technology, University of Tokyo, 7-3-1 Hongo, Tokyo 113-8656 (Japan); Urushidani, S. [National Institute of Informatics, 2-1-2 Hitotsubashi, Chiyoda-ku, Tokyo 101-8430 (Japan)

    2010-07-15

    This paper presents a proposal of the remote participation system for the international thermonuclear experimental reactor (ITER). The object of this paper is to clarify technical issues to analyze the ITER data safely and conveniently. The Japanese case is considered as an example, but technologies presented here can be used worldwide. Major technical issues are as follows: (1) the long distance data transfer; (2) the massive data server; (3) the secure network; (4) the convenient and fast data analysis system. Raw data of ITER can be transferred from France to Japan in a short time by optimizing TCP/IP parameters. The virtual private network (VPN) technology provides a secure environment of the data mirroring and the distributed computation. The analysis server with the WEB user interface enables physicists to analyze the ITER data from the Internet. Streaming data, such as plasma parameters in the steady state, video and sound of the ITER plasma and the status of experiment, which provides feeling of reality, are delivered by using the multi-cast technology. These technologies are being developed in SNET, which is a virtual laboratory for Japanese fusion community. International collaboration is required to develop a global distributed file system and a data analysis system further.

  7. Cold Vacuum Drying Instrument Air System Design Description (SYS 12)

    Energy Technology Data Exchange (ETDEWEB)

    SHAPLEY, B.J.; TRAN, Y.S.

    2000-06-05

    This system design description (SDD) addresses the instrument air (IA) system of the spent nuclear fuel (SNF). This IA system provides instrument quality air to the Cold Vacuum Drying (CVD) Facility. The IA system is a general service system that supports the operation of the heating, ventilation, and air conditioning (HVAC) system, the process equipment skids, and process instruments in the CVD Facility. The following discussion is limited to the compressor, dryer, piping, and valving that provide the IA as shown in Drawings H-1-82222, Cold Vacuum Drying Facility Mechanical Utilities Compressed & Instrument Air P&ID, and H-1.82161, Cold Vacuum Drying Facility Process Equipment Skid P&ID MCO/Cusk Interface. Figure 1-1 shows the physical location of the 1A system in the CVD Facility.

  8. The JET vacuum interspace system

    International Nuclear Information System (INIS)

    Orchard, J.; Scales, S.

    1999-01-01

    In the past JET has suffered from a number of vacuum leaks on components such as bellows, windows and feedthroughs due, in part, to the adverse conditions, including high mechanical forces, which may prevail during plasma operation. Therefore before the recent Tritium experiments on JET it was deemed prudent to manufacture and install items with a secondary containment or interspace in order to minimise the effect of failure of the primary vacuum barrier on both the leak integrity of the machine and the outcome of the experiments. This paper describes the philosophy, logistics, method and implementation of an integrated connection and monitoring system on the 330 interspaces currently in position on the JET machine. Using the JET leak database comparisons are drawn of leak failure rates of the components allied to the number of operational hours, prior to the system being present and after installation and commissioning, and the case of detection compared to the previous situation. An argument is also presented on the feasibility and adaptability of this system to any large complex machine and the benefits to be obtained in reduction of leaks and operational down time. (author)

  9. ITER fast ion collective Thomson scattering, conceptual design of 60 GHz system

    International Nuclear Information System (INIS)

    Meo, F.; Bindslev, H.; Korsholm, S.B.

    2007-08-01

    The collective Thomson scattering diagnostic for ITER at the 60 GHz range is capable of measuring the fast ion distribution parallel and perpendicular to the magnetic field at different radial locations simultaneously. The design is robust technologically with no moveable components near the plasma. The fast ion CTS diagnostic consists of two separate systems. Each system has its own RF launcher and separate set of detectors. The first system measures the perpendicular component of the fast ion velocity distribution. It consists of radially directed RF launcher and receiver, both located in the equatorial port on the low field side (LFS). This system will be referred to by the acronym LFS-BS system referring to the location of the receiver and the fact that it measures backscattered radiation. The second part of the CTS diagnostic measures the parallel component of the fast ion distribution. It consists of an RF launcher located in the mid-plane port on the LFS and a receiver mounted on the inner vacuum vessel wall that views the plasma from between two blanket modules. This system will be referred to as HFS-FS referring to the location of the receivers and that they measure forward scattered radiation. The design of both LFS-BS and HFS-FS receivers is aimed at measuring at different spatial locations simultaneously with no moveable components near the plasma. This report is a preliminary study of the hardware design and engineering constraints for this frequency range. Section 2 conceptually describes the two systems and their main components. Section 3 clarifies the impact of design parameters such as beam widths and scattering angle on the CTS measurements. With this in hand, the ITER measurement requirements are translated into constraints on the CTS system designs. An important result in this section is that systems can be designed inside these constraints. Section 4 outlines the technical feasibility and describes in more detail the design and the engineering

  10. Vacuum system for LHC

    International Nuclear Information System (INIS)

    Groebner, O.

    1995-01-01

    The Large Hadron Collider (LHC) which is planned at CERN will be housed in the tunnel of the Large Electron Positron collider (LEP) and will store two counter-rotating proton beams with energies of up to 7 TeV in a 27 km accelerator/storage ring with superconducting magnets. The vacuum system for the LHC will be at cryogenic temperatures (between 1.9 and 20 K) and will be exposed to synchrotron radiation emitted by the protons. A stringent limitation on the vacuum is given by the energy deposition in the superconducting coils of the magnets due to nuclear scattering of the protons on residual gas molecules because this may provoke a quench. This effect imposes an upper limit to a local region of increased gas density (e.g. a leak), while considerations of beam lifetime (100 h) will determine more stringent requirements on the average gas density. The proton beam creates ions from the residual gas which may strike the vacuum chamber with sufficient energy to lead to a pressure 'run-away' when the net ion induced desorption yield exceeds a stable limit. These dynamic pressure effects will be limited to an acceptable level by installing a perforated 'beam screen' which shields the cryopumped gas molecules at 1.9 K from synchrotron radiation and which also absorbs the synchrotron radiation power at a higher and, therefore, thermodynamically more efficient temperature. (author)

  11. System engineering and configuration management in ITER

    International Nuclear Information System (INIS)

    Chiocchio, S.; Martin, E.; Barabaschi, P.; Bartels, Hans Werner; How, J.; Spears, W.

    2007-01-01

    The construction of ITER will represent a major challenge for the fusion community at large, because of the intrinsic complexity of the tokamak design, the large number of different systems which are all essential for its operation, the worldwide distribution of the design activities and the unusual procurement scheme based on a combination of in-kind and directly funded deliverables. A key requirement for the success of such a large project is that a systematic approach to ensure the consistency of the design with the required performance is adopted. Also, effective project management methods, tools and working practices must be deployed to facilitate the communication and collaboration among the institutions and industries involved in the project. The authors have been involved in the definition and practical implementation of the design integration and configuration control structure inside ITER and in the system engineering process during the selection and optimization of the machine configuration. In parallel, they have assessed design, drawing and documentation management software to be used for the construction phase. Here, they describe the experience gained in recent years, explain the drivers behind the selection of the documents and drawings management systems, and illustrate the scope and issues of the configuration management activities to ensure the congruence of the design, to control and track the design changes and to manage the interfaces among the ITER systems

  12. NITSOL: A Newton iterative solver for nonlinear systems

    Energy Technology Data Exchange (ETDEWEB)

    Pernice, M. [Univ. of Utah, Salt Lake City, UT (United States); Walker, H.F. [Utah State Univ., Logan, UT (United States)

    1996-12-31

    Newton iterative methods, also known as truncated Newton methods, are implementations of Newton`s method in which the linear systems that characterize Newton steps are solved approximately using iterative linear algebra methods. Here, we outline a well-developed Newton iterative algorithm together with a Fortran implementation called NITSOL. The basic algorithm is an inexact Newton method globalized by backtracking, in which each initial trial step is determined by applying an iterative linear solver until an inexact Newton criterion is satisfied. In the implementation, the user can specify inexact Newton criteria in several ways and select an iterative linear solver from among several popular {open_quotes}transpose-free{close_quotes} Krylov subspace methods. Jacobian-vector products used by the Krylov solver can be either evaluated analytically with a user-supplied routine or approximated using finite differences of function values. A flexible interface permits a wide variety of preconditioning strategies and allows the user to define a preconditioner and optionally update it periodically. We give details of these and other features and demonstrate the performance of the implementation on a representative set of test problems.

  13. Comparison of occlusion break responses and vacuum rise times of phacoemulsification systems.

    Science.gov (United States)

    Sharif-Kashani, Pooria; Fanney, Douglas; Injev, Val

    2014-07-30

    Occlusion break surge during phacoemulsification cataract surgery can lead to potential surgical complications. The purpose of this study was to quantify occlusion break surge and vacuum rise time of current phacoemulsification systems used in cataract surgery. Occlusion break surge at vacuum pressures between 200 and 600 mmHg was assessed with the Infiniti® Vision System, the WhiteStar Signature® Phacoemulsification System, and the Centurion® Vision System using gravity-fed fluidics. Centurion Active FluidicsTM were also tested at multiple intraoperative pressure target settings. Vacuum rise time was evaluated for Infiniti, WhiteStar Signature, Centurion, and Stellaris® Vision Enhancement systems. Rise time to vacuum limits of 400 and 600 mmHg was assessed at flow rates of 30 and 60 cc/minute. Occlusion break surge was analyzed by 2-way analysis of variance. The Centurion system exhibited substantially less occlusion break surge than the other systems tested. Surge area with Centurion Active Fluidics was similar to gravity fluidics at an equivalent bottle height. At all Centurion Active Fluidics intraoperative pressure target settings tested, surge was smaller than with Infiniti and WhiteStar Signature. Infiniti had the fastest vacuum rise time and Stellaris had the slowest. No system tested reached the 600-mmHg vacuum limit. In this laboratory study, Centurion had the least occlusion break surge and similar vacuum rise times compared with the other systems tested. Reducing occlusion break surge may increase safety of phacoemulsification cataract surgery.

  14. A cryogenic system design for the international thermonuclear experimental reactor (ITER)

    International Nuclear Information System (INIS)

    Slack, D.S.

    1991-01-01

    A conceptual design for ITER was completed last year. The author developed a suitable cryogenic system for ITER as part of this conceptual design effort. An overview of the design is reported. Emphasis is on the fact that cryogenics is a mature science, and a system supporting ITER needs can be made from time-proven components without loss of efficiency or reliability. Because of the large size of the ITER cryogenic system, large numbers of compressors and expanders must be used. Very high reliability is assured by arranging these components in parallel banks where servicing of individual components can be done without interruption of operations. This and other ideas based on the author's experience with Mirror Fusion Test Facility (MFTF) operations are described. 5 refs., 3 figs

  15. Status of Conceptual Design Progress for ITER Sector Sub-assembly Tools

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung Kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Ki Hak; Robert, Shaw [ITER Organization, Paul lez Durance (France)

    2010-05-15

    The ITER (International Thermonuclear Experimental Reactor) Tokamak assembly tools are purpose-built tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. Based on the design description document prepared by the ITER organization, Korea has carried out the conceptual design of assembly tools. The 40 .deg. sector assemblies sub-assembled at assembly hall are transferred to Tokamak hall using the lifting tool operated by Tokamak main cranes. In-pit assembly tools are the purpose-built assembly tools for the completion of final sector assembly at Tokamak hall. The 40 .deg. sector sub-assembly tools are composed of the upending tool, the sector sub-assembly tool, the sector lifting tool and the vacuum vessel support and bracing tools. The process of the ITER sector sub-assembly at assembly hall and status of research and development are described in this paper. The ITER Tokamak device is composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 .deg. sectors. Each VV/TFCs/VVTS 40 .deg. sector is made up of one 40 .deg. VV, two 20 .deg. TFCs and associated VVTS segments. The 40 .deg. sectors are sub-assembled at assembly hall respectively and then 9 sectors which sub-assembled at assembly hall are finally assembled at Tokamak hall. As a basic assembly component, the assembly strategy and tools for the 40 .deg. sector sub-assembly and final assembly at inpit should be developed to satisfy the basic assembly requirements of the ITER Tokamak device. Accordingly, the purpose-built assembly tools should be designed and manufactured considering assembly plan, available space, safety, easy operation, efficient maintenance, and so on. The 40 .deg. sector assembly tools are classified into 2 groups. One group is the sub-assembly tools including upending tool, lifting tool, sub-assembly tool, VV supports and bracing tools used at assembly hall and the other group is the in

  16. 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

  17. 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.

  18. EU Developments of the ITER ECRH System

    International Nuclear Information System (INIS)

    Henderson, M.

    2006-01-01

    The electron cyclotron (EC) heating and current drive (H (and) CD) system of ITER will deliver 20 MW/CW in the plasma at 170 GHz for H (and) CD in addition to 2.5 MW/3 s at 120 GHz for plasma start-up. The EC system is composed of power supplies (PS), up to 24 H (and) CD gyrotrons (1 to 2 MW tubes), 3 start-up gyrotrons (1 MW tubes), 24 transmission lines and two sets of launching antennas: equatorial (EL) and upper (UL) launchers. Under the present ITER procurement package the EU is responsible for one third of the H (and) CD 170 GHz gyrotrons, all PSs associated with the H (and) CD system, and the whole set (4) of upper launchers. In all areas of participation, the EU EC partnership (coordinated by the European Fusion Development Association - EFDA) aims toward advancing the technology of each of these subsystems. For example, procurement of Pulse Step Modulator (PSM) HVPS is under consideration, which might have equivalent costs to the present ITER design (thyristor HVPS and HV series switch), but with an increased flexibility in operation and variation in the EC power waveform. The EU is at the forefront in gyrotron research and is developing a 2 MW CW 170 GHz coaxial cavity gyrotron offering an increase in output power while maintaining moderate power densities in the gyrotron cavity and collector. THALES R in collaboration with its EFDA partners (FZK, CRPP, TEKES) is manufacturing a series of prototype tubes in three phases of typically 1 s, 100 s and then CW pulse capacity (∼ 20 10 ). A 2 MW, CW gyrotron test facility is being built at CRPP that will be used to develop the 2 MW coaxial tube, in addition to testing various components required by the EC system. EFDA has undertaken a parallel development of two launcher options: front (FS) and remote (RS) steering, with the aim of providing an optimum launcher for ITER weighing EC physics aspects and operation reliability. The FS launcher (ITER reference design) offers a significant enhancement in physics

  19. 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.

  20. TFCX pumped limiter and vacuum pumping system design and analysis

    International Nuclear Information System (INIS)

    Haines, J.R.

    1985-04-01

    Impurity control system design and performance studies were performed in support of the Tokamak Fusion Core Experiment (TFCX) pre-conceptual design. Efforts concentrated on pumped limiter and vacuum pumping system design configuration, thermal/mechanical and erosion lifetime performance of the limiter protective surface, and helium ash removal performance. The reference limiter design forms a continuous toroidal belt at the bottom of the device and features a flat surface with a single leading edge. The vacuum pumping system features large vacuum ducts (diameter approximately 1 m) and high-speed, compound cryopumps. Analysis results indicate that the limiter/vacuum pumping system design provides adequate helium ash removal. Erosion, primarily by disruption-induced vaporization and/or melting, limits the protective surface lifetime to about one calendar year or only about 60 full-power hours of operation. In addition to evaluating impurity control system performance for nominal TFCX conditions, these studies attempt to focus on the key plasma physics and engineering design issues that should be addressed in future research and development programs

  1. ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

    International Nuclear Information System (INIS)

    Berry, Jan; Ferrada, Juan J.; Curd, Warren; Dell Orco, Giovanni; Barabash, Vladimir; Kim, Seokho H.

    2011-01-01

    During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support

  2. Apollo telescope mount thermal systems unit thermal vacuum test

    Science.gov (United States)

    Trucks, H. F.; Hueter, U.; Wise, J. H.; Bachtel, F. D.

    1971-01-01

    The Apollo Telescope Mount's thermal systems unit was utilized to conduct a full-scale thermal vacuum test to verify the thermal design and the analytical techniques used to develop the thermal mathematical models. Thermal vacuum test philosophy, test objectives configuration, test monitoring, environment simulation, vehicle test performance, and data correlation are discussed. Emphasis is placed on planning and execution of the thermal vacuum test with particular attention on problems encountered in conducting a test of this maguitude.

  3. Cold Vacuum Drying facility sanitary sewage collection system design description

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) sanitary sewage collection system. The sanitary sewage collection system provides collection and storage of effluents and raw sewage from the CVDF to support the cold vacuum drying process. This system is comprised of a sanitary sewage holding tank and pipes for collection and transport of effluents to the sanitary sewage holding tank

  4. Vacuum system design and tritium inventory for the TFTR charge exchange diagnostic

    International Nuclear Information System (INIS)

    Medley, S.S.

    1979-05-01

    The charge exchange diagnostic for the TFTR is comprised of two analyzer systems which contain a total of twenty independent mass/energy analyzers and one diagnostic neutral beam tentatively rated at 80 keV, 15 A. The associated vacuum systems were analyzed using the Vacuum System Transient Simulator (VSTS) computer program which models the transient transport of multi-gas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced cost, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus

  5. A remote joint system for large vacuum ducts

    International Nuclear Information System (INIS)

    Hagmann, D.B.; Coughlan, J.B.

    1983-01-01

    A large remote vacuum duct joining system has been developed for fusion machines that uses several two-jaw screwdriven clamps. The preferred location for clamp installation is inside the vacuum duct where access space is available for the actuating device. It also decreases space needed for handling operations exterior to the duct. The clamp system is unique in that it is low cost, applies force directly over the seal, permits leak testing to the seal annulus, is highly reliable, can be remotely replaced, and is usable on a variety of other applications

  6. Review of vacuum systems for x-ray lithography light sources

    International Nuclear Information System (INIS)

    Schuchman, J.C.

    1990-01-01

    This paper will review and give a status report on vacuum systems for X-Ray lithography light sources. It will include conventional machines and compact machines (machines using superconducting magnets). The vacuum systems will be described and compared with regard to basic machine parameters, pumping systems, types of pumps, chamber design and material, gauging and diagnostics, and machine performane. 23 refs., 8 figs., 1 tab

  7. Troubles in vacuum system and radiation exposure

    International Nuclear Information System (INIS)

    Konno, Osamu

    1978-01-01

    It is about eleven years since the LINAC of 300 MeV in Tohoku University has first accelerated electrons. The maintenance and improvement of the accelerator used more than 10 years now give the related personnel an important problem of radiation exposure. 40 days were required for the maintenance and checking-up in 1977, and other 26 days were used for other construction works. The troubles in the vacuum system occurred 81 times in total. The vacuum system is divided into two subsystems, each being provided with a leak detector. Either of them enables to detect and locate the leak. Silver-alloy brazing of a duct with a cooling water tube has deteriorated in the strength because of repeated baking temperature and/or the copper tubes for cooling have been eroded due to the large local cell action by purified water. The similar phenomena have occurred in RF windows, outside of which is cooled with water. Carbonaceous matter has stuck to the element of the ion pump, but successfully been cleaned. Though the energy compression system was installed for the efficient use of electrons, the troubles due to overheating of the current monitor have increased because of its limited space, and the change of location was made. Considerable surface residual radiation dose was found at some parts of transport system, and a few personnel have been exposed to radiation over 1000 mrem/year as a result of the troubles in vacuum system. (Wakatsuki, Y.)

  8. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  9. 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

  10. Vacuum Systems Consensus Guideline for Department of Energy Accelerator Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Casey,R.; Haas, E.; Hseuh, H-C.; Kane, S.; Lessard, E.; Sharma, S.; Collins, J.; Toter, W. F.; Olis, D. R.; Pushka, D. R.; Ladd, P.; Jobe, R. K.

    2008-09-09

    Vacuum vessels, including evacuated chambers and insulated jacketed dewars, can pose a potential hazard to equipment and personnel from collapse, rupture due to back-fill pressurization, or implosion due to vacuum window failure. It is therefore important to design and operate vacuum systems in accordance with applicable and sound engineering principles. 10 CFR 851 defines requirements for pressure systems that also apply to vacuum vessels subject to back-fill pressurization. Such vacuum vessels are potentially subject to the requirements of the American Society of Mechanical Engineers (ASME) Pressure Vessel Code Section VIII (hereafter referred to as the 'Code'). However, the scope of the Code excludes vessels with internal or external operating pressure that do not exceed 15 pounds per square inch gauge (psig). Therefore, the requirements of the Code do not apply to vacuum systems provided that adequate pressure relief assures that the maximum internal pressure within the vacuum vessel is limited to less than 15 psig from all credible pressure sources, including failure scenarios. Vacuum vessels that cannot be protected from pressurization exceeding 15 psig are subject to the requirements of the Code. 10 CFR 851, Appendix A, Part 4, Pressure Safety, Section C addresses vacuum system requirements for such cases as follows: (c) When national consensus codes are not applicable (because of pressure range, vessel geometry, use of special materials, etc.), contractors must implement measures to provide equivalent protection and ensure a level of safety greater than or equal to the level of protection afforded by the ASME or applicable state or local code. Measures must include the following: (1) Design drawings, sketches, and calculations must be reviewed and approved by a qualified independent design professional (i.e., professional engineer). Documented organizational peer review is acceptable. (2) Qualified personnel must be used to perform examinations

  11. Iterative solution of linear systems in the 20­th century

    NARCIS (Netherlands)

    Saad, Y.; Vorst, H.A. van der

    2000-01-01

    This paper sketches the main research developments in the area of iterative methods for solving linear systems during the 20th century. Although iterative methods for solving linear systems find their origin in the early nineteenth century (work by Gauss), the field has seen an explosion of

  12. Development of the vacuum system pressure responce analysis code PRAC

    International Nuclear Information System (INIS)

    Horie, Tomoyoshi; Kawasaki, Kouzou; Noshiroya, Shyoji; Koizumi, Jun-ichi.

    1985-03-01

    In this report, we show the method and numerical results of the vacuum system pressure responce analysis code. Since fusion apparatus is made up of many vacuum components, it is required to analyze pressure responce at any points of the system when vacuum system is designed or evaluated. For that purpose evaluating by theoretical solution is insufficient. Numerical analysis procedure such as finite difference method is usefull. In the PRAC code (Pressure Responce Analysis Code), pressure responce is obtained solving derivative equations which is obtained from the equilibrium relation of throughputs and contain the time derivative of pressure. As it considers both molecular and viscous flows, the coefficients of the equation depend on the pressure and the equations become non-linear. This non-linearity is treated as piece-wise linear within each time step. Verification of the code is performed for the simple problems. The agreement between numerical and theoretical solutions is good. To compare with the measured results, complicated model of gas puffing system is analyzed. The agreement is well for practical use. This code will be a useful analytical tool for designing and evaluating vacuum systems such as fusion apparatus. (author)

  13. Vacuum system for Advanced Test Accelerator

    International Nuclear Information System (INIS)

    Denhoy, B.S.

    1981-01-01

    The Advanced Test Accelerator (ATA) is a pulsed linear electron beam accelerator designed to study charged particle beam propagation. ATA is designed to produce a 10,000 amp 50 MeV, 70 ns electron beam. The electron beam acceleration is accomplished in ferrite loaded cells. Each cell is capable of maintaining a 70 ns 250 kV voltage pulse across a 1 inch gap. The electron beam is contained in a 5 inch diameter, 300 foot long tube. Cryopumps turbomolecular pumps, and mechanical pumps are used to maintain a base pressure of 2 x 10 -6 torr in the beam tube. The accelerator will be installed in an underground tunnel. Due to the radiation environment in the tunnel, the controlling and monitoring of the vacuum equipment, pressures and temperatures will be done from the control room through a computer interface. This paper describes the vacuum system design, the type of vacuum pumps specified, the reasons behind the selection of the pumps and the techniques used for computer interfacing

  14. Vacuum system for Advanced Test Accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Denhoy, B.S.

    1981-09-03

    The Advanced Test Accelerator (ATA) is a pulsed linear electron beam accelerator designed to study charged particle beam propagation. ATA is designed to produce a 10,000 amp 50 MeV, 70 ns electron beam. The electron beam acceleration is accomplished in ferrite loaded cells. Each cell is capable of maintaining a 70 ns 250 kV voltage pulse across a 1 inch gap. The electron beam is contained in a 5 inch diameter, 300 foot long tube. Cryopumps turbomolecular pumps, and mechanical pumps are used to maintain a base pressure of 2 x 10/sup -6/ torr in the beam tube. The accelerator will be installed in an underground tunnel. Due to the radiation environment in the tunnel, the controlling and monitoring of the vacuum equipment, pressures and temperatures will be done from the control room through a computer interface. This paper describes the vacuum system design, the type of vacuum pumps specified, the reasons behind the selection of the pumps and the techniques used for computer interfacing.

  15. New baking system for the RFX vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Collarin, P.; Luchetta, A.; Sonato, P.; Toigo, V.; Zaccaria, P.; Zollino, G. [Universita di Padova (Italy)

    1996-12-31

    A heating system based on eddy currents has been developed for the vacuum vessel of the RFX Reversed Field Pinch device. After a testing phase, carried out at low power, the final power supply system has been designed and installed. It has been used during last year to bake out the vessel and the graphite first wall up to 320{degree}C. Recently the heating system has been completed with a control system that allows for baking sessions with an automatic control of the vacuum vessel temperature and for pulse sessions with a heated first wall. After the description of the preliminary analyses and tests, and of the main characteristics of the power supply and control systems, the experimental results of the baking sessions performed during last year are presented. 6 refs., 7 figs.

  16. New baking system for the RFX vacuum vessel

    International Nuclear Information System (INIS)

    Collarin, P.; Luchetta, A.; Sonato, P.; Toigo, V.; Zaccaria, P.; Zollino, G.

    1996-01-01

    A heating system based on eddy currents has been developed for the vacuum vessel of the RFX Reversed Field Pinch device. After a testing phase, carried out at low power, the final power supply system has been designed and installed. It has been used during last year to bake out the vessel and the graphite first wall up to 320 degree C. Recently the heating system has been completed with a control system that allows for baking sessions with an automatic control of the vacuum vessel temperature and for pulse sessions with a heated first wall. After the description of the preliminary analyses and tests, and of the main characteristics of the power supply and control systems, the experimental results of the baking sessions performed during last year are presented. 6 refs., 7 figs

  17. LSODKR, Stiff Ordinary Differential Equations (ODE) System Solver with Krylov Iteration and Root-finding

    International Nuclear Information System (INIS)

    Hindmarsh, A.D.; Brown, P.N.

    1996-01-01

    1 - Description of program or function: LSODKR is a new initial value ODE solver for stiff and non-stiff systems. It is a variant of the LSODPK and LSODE solvers, intended mainly for large stiff systems. The main differences between LSODKR and LSODE are the following: a) for stiff systems, LSODKR uses a corrector iteration composed of Newton iteration and one of four preconditioned Krylov subspace iteration methods. The user must supply routines for the preconditioning operations, b) within the corrector iteration, LSODKR does automatic switching between functional (fix point) iteration and modified Newton iteration, c) LSODKR includes the ability to find roots of given functions of the solution during the integration. 2 - Method of solution: Integration is by Adams or BDF (Backward Differentiation Formula) methods, at user option. Corrector iteration is by Newton or fix point iteration, determined dynamically. Linear system solution is by a preconditioned Krylov iteration, selected by user from Incomplete Orthogonalization Method, Generalized Minimum Residual Method, and two variants of Preconditioned Conjugate Gradient Method. Preconditioning is to be supplied by the user. 3 - Restrictions on the complexity of the problem: None

  18. 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

  19. Status of ITER blanket attachment design and related R and D

    International Nuclear Information System (INIS)

    Sadakov, S.; Khomiakov, S.; Calcagno, B.; Chappuis, Ph.; Dellopoulos, G.; Kolganov, V.; Merola, M.; Poddubnyi, I.; Raffray, R.; Raharijaona, J.J.; Ulrickson, M.; Zhmakin, A.

    2013-01-01

    Highlights: • ITER blanket attachment system went through a significant design upgrade and become basically compliant with specified design loads and required cyclic lifetime. • Upgrade of flexible supports allowed the doubling of cross sections of central bolts. Ceramic coatings were relocated to much larger areas on conical pairs screwed into shield blocks. • Key pads were relocated from keys of vacuum vessel into keyways of shield blocks and reshaped to enlarge areas of lateral interfaces with ceramic electro-insulating coatings. • Ceramic coatings are hidden between pads and enclosures in keyways with a purpose to minimize their wear rate, which depends on peak friction stress and cyclic sliding path. • Ceramic coatings to be verified by experiment, with several R and D aimed to collect statistically sufficient data on their reliability and durability in ITER relevant cyclic loading conditions. -- Abstract: Main function of the ITER blanket system [1–3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R and D is also briefly described

  20. Iterative Learning Control design for uncertain and time-windowed systems

    NARCIS (Netherlands)

    Wijdeven, van de J.J.M.

    2008-01-01

    Iterative Learning Control (ILC) is a control strategy capable of dramatically increasing the performance of systems that perform batch repetitive tasks. This performance improvement is achieved by iteratively updating the command signal, using measured error data from previous trials, i.e., by

  1. Design of ITER-FEAT RF heating and current drive systems

    International Nuclear Information System (INIS)

    Bosia, G.; Kobayashi, N.; Ioki, K.; Bibet, P.; Koch, R.; Chavan, R.; Tran, M.Q.; Takahashi, K.; Kuzikov, S.; Vdovin, V.

    2001-01-01

    Three radio frequency (RF) heating and current drive (H and CD) systems are being designed for ITER-FEAT: an electron cyclotron (EC), an ion cyclotron (IC) and a lower hybrid (LH) System. The launchers of the RF systems use four ITER equatorial ports and are fully interchangeable. They feature equal power outputs (20 MW/port), similar neutron shielding performance, and identical interfaces with the other machine components. An outline of the design is given in the paper. (author)

  2. Numerical simulation on bake-out of the ITER diagnostic upper port plug

    International Nuclear Information System (INIS)

    Pak, S.; Pitcher, C.S.; Kalish, M.R.; Cheon, M.S.; Seon, C.R.; Lee, H.G.

    2010-01-01

    The diagnostic upper port plug in ITER is fixed to the upper port of the vacuum vessel as a cantilevered beam with bolts and forms a primary vacuum boundary. It needs to be baked out for outgassing before normal operation. This study calculated the required bake-out time and the transient thermal stress during baking for the diagnostic upper port plug. The calculation was done through numerical simulation. The analysis took into consideration the gradual temperature increase of working fluid. In order to look into the effect of radiation heat transfer from the upper port plug to the vacuum vessel port, the upper vacuum vessel port was included in this analysis.

  3. Implementation of EPICS based vacuum control system for variable energy cyclotron centre, Kolkata

    Science.gov (United States)

    Roy, Anindya; Bhole, R. B.; Nandy, Partha P.; Yadav, R. C.; Pal, Sarbajit; Roy, Amitava

    2015-03-01

    The vacuum system of the Room Temperature (K = 130) Cyclotron of Variable Energy Cyclotron Centre is comprised of vacuum systems of main machine and Beam Transport System. The vacuum control system is upgraded to a PLC based Automated system from the initial relay based Manual system. The supervisory control of the vacuum system is implemented in Experimental Physics and Industrial Control System (EPICS). An EPICS embedded ARM based vacuum gauge controller is developed to mitigate the requirement of vendor specific gauge controller for gauges and also for seamless integration of the gauge controllers with the control system. A set of MS-Windows ActiveX components with embedded EPICS Channel Access interface are developed to build operator interfaces with less complex programming and to incorporate typical Windows feature, e.g., user authentication, file handling, better fonts, colors, mouse actions etc. into the operator interfaces. The control parameters, monitoring parameters, and system interlocks of the system are archived in MySQL based EPICS MySQL Archiver developed indigenously. In this paper, we describe the architecture, the implementation details, and the performance of the system.

  4. Implementation of EPICS based vacuum control system for variable energy cyclotron centre, Kolkata

    Energy Technology Data Exchange (ETDEWEB)

    Roy, Anindya, E-mail: r-ani@vecc.gov.in; Bhole, R. B.; Nandy, Partha P.; Yadav, R. C.; Pal, Sarbajit; Roy, Amitava [Variable Energy Cyclotron Centre, 1/AF Bidhan Nagar, Kolkata 700064 (India)

    2015-03-15

    The vacuum system of the Room Temperature (K = 130) Cyclotron of Variable Energy Cyclotron Centre is comprised of vacuum systems of main machine and Beam Transport System. The vacuum control system is upgraded to a PLC based Automated system from the initial relay based Manual system. The supervisory control of the vacuum system is implemented in Experimental Physics and Industrial Control System (EPICS). An EPICS embedded ARM based vacuum gauge controller is developed to mitigate the requirement of vendor specific gauge controller for gauges and also for seamless integration of the gauge controllers with the control system. A set of MS-Windows ActiveX components with embedded EPICS Channel Access interface are developed to build operator interfaces with less complex programming and to incorporate typical Windows feature, e.g., user authentication, file handling, better fonts, colors, mouse actions etc. into the operator interfaces. The control parameters, monitoring parameters, and system interlocks of the system are archived in MySQL based EPICS MySQL Archiver developed indigenously. In this paper, we describe the architecture, the implementation details, and the performance of the system.

  5. Implementation of EPICS based vacuum control system for variable energy cyclotron centre, Kolkata

    International Nuclear Information System (INIS)

    Roy, Anindya; Bhole, R. B.; Nandy, Partha P.; Yadav, R. C.; Pal, Sarbajit; Roy, Amitava

    2015-01-01

    The vacuum system of the Room Temperature (K = 130) Cyclotron of Variable Energy Cyclotron Centre is comprised of vacuum systems of main machine and Beam Transport System. The vacuum control system is upgraded to a PLC based Automated system from the initial relay based Manual system. The supervisory control of the vacuum system is implemented in Experimental Physics and Industrial Control System (EPICS). An EPICS embedded ARM based vacuum gauge controller is developed to mitigate the requirement of vendor specific gauge controller for gauges and also for seamless integration of the gauge controllers with the control system. A set of MS-Windows ActiveX components with embedded EPICS Channel Access interface are developed to build operator interfaces with less complex programming and to incorporate typical Windows feature, e.g., user authentication, file handling, better fonts, colors, mouse actions etc. into the operator interfaces. The control parameters, monitoring parameters, and system interlocks of the system are archived in MySQL based EPICS MySQL Archiver developed indigenously. In this paper, we describe the architecture, the implementation details, and the performance of the system

  6. Disruptions in ITER and strategies for their control and mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Lehnen, M., E-mail: michael.lehnen@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Aleynikova, K.; Aleynikov, P.B.; Campbell, D.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Drewelow, P. [Max-Planck-Institut für Plasmaphysik, Greifswald branch, EURATOM Ass., D-17491 Greifswald (Germany); Eidietis, N.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Gasparyan, Yu. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoe sh. 31, Moscow 115409 (Russian Federation); Granetz, R.S. [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Gribov, Y. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Hartmann, N. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research—Plasma Physics, Association EURATOM-FZJ, Trilateral Euregio Cluster, 52425 Jülich (Germany); Hollmann, E.M. [University of California-San Diego, La Jolla, CA 92093 (United States); Izzo, V.A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Jachmich, S. [Laboratory for Plasma Physics, ERM/KMS, Association EURATOM – Belgian State, B-1000 Brussels (Belgium); Kim, S.-H.; Kočan, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Koslowski, H.R. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research—Plasma Physics, Association EURATOM-FZJ, Trilateral Euregio Cluster, 52425 Jülich (Germany); Kovalenko, D. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Kruezi, U. [CCFE, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); and others

    2015-08-15

    The thermal and electromagnetic loads related to disruptions in ITER are substantial and require careful design of tokamak components to ensure they reach the projected lifetime and to ensure that safety relevant components fulfil their function for the worst foreseen scenarios. The disruption load specifications are the basis for the design process of components like the full-W divertor, the blanket modules and the vacuum vessel and will set the boundary conditions for ITER operations. This paper will give a brief overview on the disruption loads and mitigation strategies for ITER and will discuss the physics basis which is continuously refined through the current disruption R&D programs.

  7. Systems and methods for analyzing liquids under vacuum

    Science.gov (United States)

    Yu, Xiao-Ying; Yang, Li; Cowin, James P.; Iedema, Martin J.; Zhu, Zihua

    2013-10-15

    Systems and methods for supporting a liquid against a vacuum pressure in a chamber can enable analysis of the liquid surface using vacuum-based chemical analysis instruments. No electrical or fluid connections are required to pass through the chamber walls. The systems can include a reservoir, a pump, and a liquid flow path. The reservoir contains a liquid-phase sample. The pump drives flow of the sample from the reservoir, through the liquid flow path, and back to the reservoir. The flow of the sample is not substantially driven by a differential between pressures inside and outside of the liquid flow path. An aperture in the liquid flow path exposes a stable portion of the liquid-phase sample to the vacuum pressure within the chamber. The radius, or size, of the aperture is less than or equal to a critical value required to support a meniscus of the liquid-phase sample by surface tension.

  8. Comparison of cold and warm vacuum systems for intersecting storage rings

    International Nuclear Information System (INIS)

    Halama, H.J.; Herrera, J.C.

    1975-01-01

    In storage rings employing superconducting magnets, the use of a cold bore as a cryopump appears, at first glance, as simple and economical. Since the selection of a cold or warm vacuum system has far-reaching implications on the basic design, each system is considered in some detail. The theoretical and practical limitations imposed on the maximum beam current by the gas desorption from the chamber walls are discussed. A realistic design of a cold vacuum chamber is developed and then compared with the proposed warm ISABELLE vacuum system. The comparison shows that the warm approach is preferable. (U.S.)

  9. Vacuum Technology

    Energy Technology Data Exchange (ETDEWEB)

    Biltoft, P J

    2004-10-15

    The environmental condition called vacuum is created any time the pressure of a gas is reduced compared to atmospheric pressure. On earth we typically create a vacuum by connecting a pump capable of moving gas to a relatively leak free vessel. Through operation of the gas pump the number of gas molecules per unit volume is decreased within the vessel. As soon as one creates a vacuum natural forces (in this case entropy) work to restore equilibrium pressure; the practical effect of this is that gas molecules attempt to enter the evacuated space by any means possible. It is useful to think of vacuum in terms of a gas at a pressure below atmospheric pressure. In even the best vacuum vessels ever created there are approximately 3,500,000 molecules of gas per cubic meter of volume remaining inside the vessel. The lowest pressure environment known is in interstellar space where there are approximately four molecules of gas per cubic meter. Researchers are currently developing vacuum technology components (pumps, gauges, valves, etc.) using micro electro mechanical systems (MEMS) technology. Miniature vacuum components and systems will open the possibility for significant savings in energy cost and will open the doors to advances in electronics, manufacturing and semiconductor fabrication. In conclusion, an understanding of the basic principles of vacuum technology as presented in this summary is essential for the successful execution of all projects that involve vacuum technology. Using the principles described above, a practitioner of vacuum technology can design a vacuum system that will achieve the project requirements.

  10. Vacuum system for the test accumulation ring for the NUMATRON project (TARN)

    International Nuclear Information System (INIS)

    Mizobuchi, Akira

    1979-01-01

    The vacuum system for the test accumulation ring for the NUMATRON project (TARN) being constructed in the Institute for Nuclear Study, University of Tokyo, for the purpose of accumulating fundamental data on high energy, heavy ion acceleration is presented. The main specifications of TARN are as follows: repetition rate 1 Hz, average radius 5.06 m, the radius of curvature 1.333 m, circumference 31.795 m, the length of a unit cell 3.974 m, the length of a long straight section 1.8 m, the number of unit cells 8, superperiodicity 8, and the structure of a unit cell FODO. The vacuum level of TARN lower than 1 x 10 -10 Torr is necessary. The preliminary test apparatuses, number one and two, for obtaining the basic data on superhigh vacuum were constructed. The characteristics of a molecular vacuum pump and the high temperature baking effect were tested with the number one test apparatus. The glow discharge cleaning effect of the model vacuum chamber for deflecting magnets was tested, utilizing the number two test apparatus. The pressure curve in course of time during pump down, the residual gas components, the residual gas spectrum in vacuum state and the change of residual gas components before and after the glow discharge cleaning are shown as these testing results. The vacuum gauges, which are used for TARN for measuring both the total pressure and the partial pressure of gas, are explained. The vacuum system, the vacuum chamber, the vacuum pumps, the expected exhausting speed and the partial functional test results for the vacuum system are introduced. (Nakai, Y.)

  11. Preliminary test of vacuum system of SPring-8 synchrotron

    International Nuclear Information System (INIS)

    Otsuka, Hideo; Yoneyama, Katsuji; Shimada, Taihei

    1992-01-01

    In this report, on the several points which it is desirable to be confirmed beforehand when the vacuum system of the synchrotron is designed, the results of the test by making the vacuum system for one normal cell simulating the actual machine are reported. Further by making three types of sample ducts, the ordinary type of 1.65 mm thickness, the rib-reinforced type of 0.3 mm thickness and bellows structure, the rate of released gas and the effect to eddy current were examined, for the purpose of taking the countermeasures to the eddy current accompanying the synchrotron operation. The mockup of the normal cell was composed of a duct of 10 m length, two ion pumps of 20 l/s, and a turbo molecular pump of 50 l/s, and the material of the ducts is SUS316L, and the sectional form is elliptic of 72 x 34. In the vacuum test of the normal cell, the exhaust characteristics at the time of rise and pressure distribution were verified. The measurement of the rate of released gas in the ducts and the measurement of current and voltage induced in the ducts are reported. The sufficient performance can be expected with the exhaust system used, and the time required for the rise of the vacuum system was clarified. (K.I.)

  12. Analysis of an innovative solar water desalination system using gravity induced vacuum

    International Nuclear Information System (INIS)

    Ayhan, T.; Al-Madani, H.

    2007-01-01

    This study presents the theoretical analysis, design and appropriate models of a new desalination system using gravity induced vacuum. The system utilizes natural means (gravity and atmospheric pressure) to create a vacuum under which water can be rapidly evaporated at much lower temperatures with less energy than conventional techniques. This technique is developed to overcome water storage, in the areas where good solar radiation (or waste heat sources) and sea water (or waste water sources). The developed system consists of an evaporator connected to condenser by means of a vacuum tank. The vapour produced in the evaporator is driven to condenser through the vacuum tank, where it condenses and collected as a product. Vacuum equivalent to 7 kPa (abs) or less can be created depending on ambient temperature of Bahrain climatic conditions. The effect of various operating conditions, namely water levels in condensation and evaporating columns on the system performance were studied. The theoretical analysis and preliminary experimental results show that the performance of this system depends on the condensation temperature

  13. Progress in the integration of the ITER plant systems in auxiliary buildings

    International Nuclear Information System (INIS)

    Kotamäki, M.; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

    2016-01-01

    Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

  14. Progress in the integration of the ITER plant systems in auxiliary buildings

    Energy Technology Data Exchange (ETDEWEB)

    Kotamäki, M., E-mail: miikka.kotamaki@iter.org; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

    2016-11-01

    Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

  15. 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.

  16. 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)

  17. Verifying large modular systems using iterative abstraction refinement

    International Nuclear Information System (INIS)

    Lahtinen, Jussi; Kuismin, Tuomas; Heljanko, Keijo

    2015-01-01

    Digital instrumentation and control (I&C) systems are increasingly used in the nuclear engineering domain. The exhaustive verification of these systems is challenging, and the usual verification methods such as testing and simulation are typically insufficient. Model checking is a formal method that is able to exhaustively analyse the behaviour of a model against a formally written specification. If the model checking tool detects a violation of the specification, it will give out a counter-example that demonstrates how the specification is violated in the system. Unfortunately, sometimes real life system designs are too big to be directly analysed by traditional model checking techniques. We have developed an iterative technique for model checking large modular systems. The technique uses abstraction based over-approximations of the model behaviour, combined with iterative refinement. The main contribution of the work is the concrete abstraction refinement technique based on the modular structure of the model, the dependency graph of the model, and a refinement sampling heuristic similar to delta debugging. The technique is geared towards proving properties, and outperforms BDD-based model checking, the k-induction technique, and the property directed reachability algorithm (PDR) in our experiments. - Highlights: • We have developed an iterative technique for model checking large modular systems. • The technique uses BDD-based model checking, k-induction, and PDR in parallel. • We have tested our algorithm by verifying two models with it. • The technique outperforms classical model checking methods in our experiments

  18. Distributed remote temperature monitoring system for INDUS-2 vacuum chambers

    International Nuclear Information System (INIS)

    Bhange, N.J.; Gothwal, P.; Fatnani, P.; Shukla, S.K.

    2011-01-01

    Indus-2, a 2.5 GeV Synchrotron Radiation Source (SRS) at Indore has a large vacuum system. The vacuum envelope of Indus-2 ring comprises of 16 dipole chambers as vital parts. Each chamber has 4 photon absorbers and three beam line ports blanked with end flanges. Temperature monitoring of critical vacuum components during operation of Indus-2 ring is an important requirement. The paper discusses a distributed, 160 channel remote temperature monitoring system developed and deployed for this purpose using microcontroller based, modular Temperature Monitoring Units (TMU). The cabling has been extensively minimized using RS485 system and keeping trip relay contacts of all units in series. For ensuring proper signal conditioning of thermocouple outputs (K-type) and successful operation over RS485 bus, many precautions were taken considering the close proximity to the storage ring. We also discuss the software for vacuum chamber temperature monitoring and safety system. The software developed using LabVIEW, has important features like modularity, client-server architecture, local and global database logging, alarms and trips, event and error logging, provision of various important configurations, communications handling etc. (author)

  19. 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.)

  20. 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

  1. Control systems for ITER diagnostics, heating and current drive

    Energy Technology Data Exchange (ETDEWEB)

    Simrock, Stefan, E-mail: stefan.simrock@iter.org

    2016-11-15

    The ITER Diagnostic, Heating and Current Drive systems might appear, on the face of it, to have very different control requirements. There are approximately 45 diagnostic systems, including magnetic sensors for plasma position and shape determination, imaging systems in the IR and visible, Thompson scattering for electron temperature and density, neutron detectors and collective scattering for alpha particle density and energy distribution. The H&CD systems encompass Electron Cyclotron Heating, using 24 1MW, 170 GHz gyrotrons and 5 steerable launchers to deliver 20 MW to the plasma, Ion Cyclotron Heating, using 8 3MW, 40–55 MHz sources and two multi-element launchers to deliver 20 MW to the plasma, and 2 Negative Ion Neutral Beam Injectors, each of which can deliver up to 16.5 MW of 1 MeV beams to the plasma. Although there are substantial differences in the needs for protection, when handling multi-MW heating systems, and in data throughput for many diagnostics, the formal processes needed to translate system requirements into Instrumentation and Control are identical. Due to the distributed procurement of ITER sub-systems and the need to integrate as painlessly as possible to CODAC, the formal processes, together with a substantial degree of standardization, are even more than usually essential. Starting from the technical, safety and protection, integration and operation requirements, a loop of functional analysis and signal listing is used to generate the controller configuration and the conceptual architecture. These elements in their turn lead to the physical and software design. The paper will describe the formal processes of control system design and the methods used by the ITER project to achieve the standardization of systems engineering practices. These have been applied to several use-cases covering all operation relevant phases such as plasma operation, maintenance, testing and conditioning. There are a number of running contracts that are developing

  2. Measurement and control system for the ITER remote handling mock-up test

    International Nuclear Information System (INIS)

    Oka, K.; Kakudate, S.; Takiguchi, Y.; Ako, K.; Taguchi, K.; Tada, E.; Ozaki, F.; Shibanuma, K.

    1998-01-01

    The mock-up test platforms composed of full-scale remote handling (RH) equipment were developed for demonstrating remote replacement of the ITER blanket and divertor. In parallel, the measurement and control system for operating these RH equipment were constructed on the basis of open architecture with object oriented feature, aiming at realization of fully-remoted automatic operation required for ITER. This paper describes the design concept of the measurement and control system for the remote handling equipment of ITER, and outlines the measured performances of the fabricated measurement system for the remote handling mock-up tests, which includes Data Acquisition System (DAS), Visual Monitoring System (VMS) and Virtual Reality System (VRS). (authors)

  3. 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)

  4. Final design of ITER port plug test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cerisier, Thierry, E-mail: thierry.cerisier@yahoo.fr [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Levesy, Bruno [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Romannikov, Alexander [Institution “Project Center ITER”, Kurchatov sq. 1, Building 3, Moscow 123182 (Russian Federation); Rumyantsev, Yuri [JSC “Cryogenmash”, Moscow reg., Balashikha 143907 (Russian Federation); Cordier, Jean-Jacques; Dammann, Alexis [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Minakov, Victor; Rosales, Natalya; Mitrofanova, Elena [JSC “Cryogenmash”, Moscow reg., Balashikha 143907 (Russian Federation); Portone, Sergey; Mironova, Ekaterina [Institution “Project Center ITER”, Kurchatov sq. 1, Building 3, Moscow 123182 (Russian Federation)

    2016-11-01

    Highlights: • We introduce the port plug test facility (purpose and status of the design). • We present the PPTF sub-systems. • We present the environmental and functional tests. • We present the occupational and nuclear safety functions. • We conclude on the achievements and next steps. - Abstract: To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill this requirement, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment. The ITER port plug test facility (PPTF) is composed of several test stands that can be used to test the port plugs whereas at the end of manufacturing (in a non-nuclear environment), or after refurbishment in the ITER hot cell facility. The PPTF provides the possibility to perform environmental (leak tightness, vacuum and thermo-hydraulic performances) and functional tests (radio frequency acceptance tests, behavior of the plugs’ steering mechanism and calibration of diagnostics) on upper and equatorial port plugs. The final design of the port plug test facility is described. The configuration of the standalone test stands and the integration in the hot cell facility are presented.

  5. Application of programmable controllers to vacuum system interlocks

    International Nuclear Information System (INIS)

    Lee, G.; Moore, D.

    1979-11-01

    This paper describes the Doublet III Vacuum Control System in which all input signals and output loads are connected to a programmable controller (PC) for logical interfacing. Input signals derived from CAMAC, control panels, limit switches, etc., are implemented as output signals to CAMAC, vacuum valves, pump motors, etc., according to a logic program stored in the PC memory. The memory can be easily programmed by anyone familar with either Boolean algebra or relay-ladder network diagrams. The program data is entered with the aid of a calculator like, keyboard instrument with LED readout displays. The PC system contains a 1024 word RAM memory with a battery backup system to provide 72 hours protection of contents in case of power failure

  6. Vacuum system for the tokamak fusion test reactor (TFTR)

    International Nuclear Information System (INIS)

    Lange, W.J.; Green, D.; Sink, D.A.

    1976-01-01

    The vacuum system for TFTR is described. Insofar as possible, conventional and ultrahigh vacuum (UHV) components and technology will be employed. Subassemblies will be prebaked in vacuum to reduce subsequent outgassing, and assembly will employ TIG welding and metal gaskets. It is not anticipated that the totally assembled torus with its numerous diagnostic appendages will be baked in situ to a high temperature, however a lower bakeout temperature (approximately 250 0 C) is under consideration. Final vacuum conditioning will be performed using discharge cleaning to obtain a specific outgassing rate of less than or = to 10 -10 Torr liter/sec cm 2 hydrogen isotopes and less than or = to 10 -12 Torr liter/sec cm 2 of other gases, and a base pressure of less than or = to 5 x 10 -8 Torr

  7. CEA engineering studies and integration of the ITER diagnostic port plugs

    International Nuclear Information System (INIS)

    Doceul, L.; Walker, C.; Ingesson, C.; Ciattaglia, E.; Chappuis, P.; Portafaix, C.; Salasca, S.; Thomas, E.; Tremblay, G.; Bruyere, C.

    2007-01-01

    Most of the ITER diagnostic system is integrated in port plugs, which are water cooled stainless steel structures inserted into the vacuum-vessel ports. The port plug must perform basic functions such as providing neutron and gamma shielding, supporting the first wall armour and shielding blanket material, closing the vacuum vessel ports, while supporting the diagnostic equipment. CEA has contributed to the engineering activities on the port plugs and has more particularly focused on the design and diagnostic integration in the representative equatorial port plug Eq no. 01. The specific CEA contributions have been the engineering, structural and thermal analysis. These detailed analyses have highlighted some design issues which were worked out through different solutions. This paper contains a description of the engineering activities performed such as: the conceptual design of the Eq no. 01 port plug, the static mechanical calculations, the dynamic calculation to estimate the dynamic amplification factor due to the resonance phenomenon, the thermal assessment under the neutronic load and the seismic response of the port plug inside the vacuum vessel

  8. CEA engineering studies and integration of the ITER diagnostic port plugs

    Energy Technology Data Exchange (ETDEWEB)

    Doceul, L. [Association Euratom-CEA sur la Fusion Controlee, Centre d' Etudes de Cadarache, F-13108 Saint-Paul-Lez-Durance Cedex (France)], E-mail: louis.doceul@cea.fr; Walker, C. [ITER International Team, Boltzmannstr. 2, D-85748 Garching bei Muenchen (Germany); Ingesson, C.; Ciattaglia, E. [EFDA CSU - Garching, Boltzmannstr. 2, D-85748 Garching bei Muenchen (Germany); Chappuis, P.; Portafaix, C.; Salasca, S.; Thomas, E.; Tremblay, G.; Bruyere, C. [Association Euratom-CEA sur la Fusion Controlee, Centre d' Etudes de Cadarache, F-13108 Saint-Paul-Lez-Durance Cedex (France)

    2007-10-15

    Most of the ITER diagnostic system is integrated in port plugs, which are water cooled stainless steel structures inserted into the vacuum-vessel ports. The port plug must perform basic functions such as providing neutron and gamma shielding, supporting the first wall armour and shielding blanket material, closing the vacuum vessel ports, while supporting the diagnostic equipment. CEA has contributed to the engineering activities on the port plugs and has more particularly focused on the design and diagnostic integration in the representative equatorial port plug Eq no. 01. The specific CEA contributions have been the engineering, structural and thermal analysis. These detailed analyses have highlighted some design issues which were worked out through different solutions. This paper contains a description of the engineering activities performed such as: the conceptual design of the Eq no. 01 port plug, the static mechanical calculations, the dynamic calculation to estimate the dynamic amplification factor due to the resonance phenomenon, the thermal assessment under the neutronic load and the seismic response of the port plug inside the vacuum vessel.

  9. On differential operators generating iterative systems of linear ODEs of maximal symmetry algebra

    Science.gov (United States)

    Ndogmo, J. C.

    2017-06-01

    Although every iterative scalar linear ordinary differential equation is of maximal symmetry algebra, the situation is different and far more complex for systems of linear ordinary differential equations, and an iterative system of linear equations need not be of maximal symmetry algebra. We illustrate these facts by examples and derive families of vector differential operators whose iterations are all linear systems of equations of maximal symmetry algebra. Some consequences of these results are also discussed.

  10. System Configuration Management Implementation Procedure for the Cold Vacuum Drying Facility Monitoring and Control System

    International Nuclear Information System (INIS)

    ANGLESEY, M.O.

    2000-01-01

    The purpose of this document is to establish the System Configuration Management Implementation Procedure (SCMIP) for the Cold Vacuum Drying Facility (CVDF) Monitoring and Control System (MCS). This procedure provides configuration management for the process control system. The process control system consists of equipment hardware and software that controls and monitors the instrumentation and equipment associated with the CVDF processes. Refer to SNF-3090, Cold Vacuum Drying Facility Monitoring and Control System Design Description, HNF-3553, Annex B, Safety Analysis Report for the Cold Vacuum Drying Facility, and AP-CM-6-037-00, SNF Project Process Automation Software and Equipment Configuration. This SCMIP identifies and defines the system configuration items in the control system, provides configuration control throughout the system life cycle, provides configuration status accounting, physical protection and control, and verifies the completeness and correctness of these items

  11. Tritium module for ITER/Tiber system code

    International Nuclear Information System (INIS)

    Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

    1988-01-01

    A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs

  12. Numerical simulation of draining and drying procedure for the ITER Generic Equatorial Port Plug cooling system

    International Nuclear Information System (INIS)

    Tanchuk, Victor; Grigoriev, Sergey; Lyublin, Boris; Maquet, Philippe; Senik, Konstantin; Pak, Sunil; Udintsev, Victor

    2016-01-01

    Highlights: • The cooling system of the ITER Generic Equatorial Port Plug (GEPP) is of a complicated combination of horizontal and vertical channels. • The calculation model for the entire GEPP cooling circuit comprising 12 sub-circuits and built up of 2421 finite-volume elements has been developed. • Transient analysis of this model simulating the draining procedure by the KORSAR/B1 code has been performed. • Water in amount of 263 g of initial 531 kg in the GEPP remains in the dead-ends of the DSM and DFW channels in 150 s of draining procedure. • Almost 3 h are required to boil off 263 g of water trapped in the dead-ends. - Abstract: For effective vacuum leak testing all cooling circuits serving the ITER vessel and in-vessel components shall be drained and dried so that after this procedure taking less than 100 h the purge gas passing through a component has water content less than 100 ppm. This process is four-stage, with the first stage using a short blast of compressed nitrogen to blow most of water in the coolant channels out of the circuit. This process is hindered by volumes which trap water due to gravity. To remove the trapped water, it is necessary, first, to heat up the structure by hot and compressed nitrogen, and then water is evaporated by depressurized nitrogen. The cooling system of the ITER Diagnostic Equatorial Port Plugs is of a complicated hydraulic configuration. The system branching might make difficult removal of water from the piping in the scheduled draining mode. The authors have proposed the KORSAR computation code to simulate draining of the GEPP cooling circuit. The numerical simulation performed has made it possible to describe the process dynamics during draining of the entire GEPP cooling circuit and to define the process time, amount and location of residual water and evolution of two-phase flow regime.

  13. Numerical simulation of draining and drying procedure for the ITER Generic Equatorial Port Plug cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Tanchuk, Victor, E-mail: Victor.Tanchuk@sintez.niiefa.spb.su [JSC “D.V. Efremov Institute of Electrophysical Apparatus”, 196641 St. Petersburg (Russian Federation); Grigoriev, Sergey; Lyublin, Boris [JSC “D.V. Efremov Institute of Electrophysical Apparatus”, 196641 St. Petersburg (Russian Federation); Maquet, Philippe [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Senik, Konstantin [JSC “D.V. Efremov Institute of Electrophysical Apparatus”, 196641 St. Petersburg (Russian Federation); Pak, Sunil [National Fusion Research Institute, Daejeon (Korea, Republic of); Udintsev, Victor [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

    2016-11-01

    Highlights: • The cooling system of the ITER Generic Equatorial Port Plug (GEPP) is of a complicated combination of horizontal and vertical channels. • The calculation model for the entire GEPP cooling circuit comprising 12 sub-circuits and built up of 2421 finite-volume elements has been developed. • Transient analysis of this model simulating the draining procedure by the KORSAR/B1 code has been performed. • Water in amount of 263 g of initial 531 kg in the GEPP remains in the dead-ends of the DSM and DFW channels in 150 s of draining procedure. • Almost 3 h are required to boil off 263 g of water trapped in the dead-ends. - Abstract: For effective vacuum leak testing all cooling circuits serving the ITER vessel and in-vessel components shall be drained and dried so that after this procedure taking less than 100 h the purge gas passing through a component has water content less than 100 ppm. This process is four-stage, with the first stage using a short blast of compressed nitrogen to blow most of water in the coolant channels out of the circuit. This process is hindered by volumes which trap water due to gravity. To remove the trapped water, it is necessary, first, to heat up the structure by hot and compressed nitrogen, and then water is evaporated by depressurized nitrogen. The cooling system of the ITER Diagnostic Equatorial Port Plugs is of a complicated hydraulic configuration. The system branching might make difficult removal of water from the piping in the scheduled draining mode. The authors have proposed the KORSAR computation code to simulate draining of the GEPP cooling circuit. The numerical simulation performed has made it possible to describe the process dynamics during draining of the entire GEPP cooling circuit and to define the process time, amount and location of residual water and evolution of two-phase flow regime.

  14. Vacuum system design for the PEP-II B Factory High-Energy Ring

    International Nuclear Information System (INIS)

    Perkins, C.; Bostic, D.; Daly, E.

    1994-06-01

    The design of the vacuum system for the PEP-II B Factory High-Energy Ring is reviewed. The thermal design and vacuum requirements are particularly challenging in PEP-II due to high stored beam currents up to 3.0 amps in 1658 bunches. The vacuum chambers for the HER arcs are fabricated by electron beam welding extruded copper sections up to 6 m long. Design of these chambers and the vacuum PumPing configuration is described with results from vacuum and thermal analyses

  15. Assessment of CONTAIN and MELCOR for performing LOCA and LOVA analyses in ITER

    International Nuclear Information System (INIS)

    Merrill, B.J.; Hagrman, D.L.; Gaeta, M.J.; Petti, D.A.

    1994-09-01

    This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded that the MELCOR code should be the preferred code for ITER severe accident thermal hydraulic analyses. This code will require the least modification to be appropriate for calculating thermal hydraulic behavior in ITER relevant conditions that include vacuum, cryogenics, ITER temperatures, and the presence of a liquid metal test module. The assessment of the aerosol transport models in these codes concludes that several modifications would have to be made to CONTAIN and/or MELCOR to make them applicable to the aerosol transport part of severe accident analysis in ITER

  16. Seismic Design of ITER Component Cooling Water System-1 Piping

    Science.gov (United States)

    Singh, Aditya P.; Jadhav, Mahesh; Sharma, Lalit K.; Gupta, Dinesh K.; Patel, Nirav; Ranjan, Rakesh; Gohil, Guman; Patel, Hiren; Dangi, Jinendra; Kumar, Mohit; Kumar, A. G. A.

    2017-04-01

    The successful performance of ITER machine very much depends upon the effective removal of heat from the in-vessel components and other auxiliary systems during Tokamak operation. This objective will be accomplished by the design of an effective Cooling Water System (CWS). The optimized piping layout design is an important element in CWS design and is one of the major design challenges owing to the factors of large thermal expansion and seismic accelerations; considering safety, accessibility and maintainability aspects. An important sub-system of ITER CWS, Component Cooling Water System-1 (CCWS-1) has very large diameter of pipes up to DN1600 with many intersections to fulfill the process flow requirements of clients for heat removal. Pipe intersection is the weakest link in the layout due to high stress intensification factor. CCWS-1 piping up to secondary confinement isolation valves as well as in-between these isolation valves need to survive a Seismic Level-2 (SL-2) earthquake during the Tokamak operation period to ensure structural stability of the system in the Safe Shutdown Earthquake (SSE) event. This paper presents the design, qualification and optimization of layout of ITER CCWS-1 loop to withstand SSE event combined with sustained and thermal loads as per the load combinations defined by ITER and allowable limits as per ASME B31.3, This paper also highlights the Modal and Response Spectrum Analyses done to find out the natural frequency and system behavior during the seismic event.

  17. PTC-6 vacuum system: WallWaler trademark Blastrac reg-sign shot blast cleaning system

    International Nuclear Information System (INIS)

    1998-02-01

    The LTC Americas, Inc. wall decontamination technology consisted of two pneumatic hand-held tools: (1) a roto-peen scaler that used star cutters and (2) a 3-piston hammer with reciprocating bits. The hand-held tools were used in conjunction with the LTC PTC-6 vacuum system which captured dust and debris as the wall decontamination took place. Recommendations for improved worker safety and health during use of the PTC-6 vacuum system with hand-held tools include: (1) keeping all hoses and lines as orderly as possible in compliance with good housekeeping requirements; (2) ergonomic training to include techniques in lifting, bending, stooping, twisting, etc.; (3) use of a clamping system to hold hoses to the vacuum system; (4) a safety line on the air line connections; (5) use of a mechanical lifting system for waste drum removal; and (6) the use of ergonomically designed tools

  18. TFTR ultrahigh-vacuum pumping system incorporating mercury diffusion pumps

    International Nuclear Information System (INIS)

    Sink, D.A.; Sniderman, M.

    1976-06-01

    The TFTR vacuum vessel will have a system of four 61 cm diameter mercury diffusion pumps to provide a base pressure in the 10 -8 to 10 -9 Torr range as well as a low impurity level within the vessel. The system, called the Torus Vacuum Pumping System (TVPS), will be employed with the aid of an occasional 250 0 C bakeout in situ as well as periodic applications of aggressive discharge cleaning. The TVPS is an ultrahigh-vacuum (UHV) system using no elastomers as well as being a closed system with respect to tritium or any tritiated gases. The backing system employing approximately 75 all-metal isolation valves is designed with the features of redundancy and flexibility employed in a variety of ways to meet the fundamental requirements and functions enumerated for the TVPS. Since the design, is one which is a modification of the conceptual design of the TVPS, those features which have changed are discussed. Calculations are presented for the major performance parameters anticipated for the TVPS and include conductances, effective pumping speeds, base pressures, operating parameters, getter pump parameters, and calculations of time constants associated with leak checking. Modifications in the vacuum pumping system for the guard regions on the twelve bellows sections are presented so that it is compatible with the main TVPS. The bellows pumping system consists of a mechanical pump unit, a zirconium aluminum getter pump unit and a residual gas analyzer. The control and management of the TVPS is described with particular attention given to providing both manual and automatic control at a local station and at the TFTR Central Control. Such operations as testing, maintenance, leak checking, startup, bakeout, and various other operations are considered in some detail. Various aspects related to normal pulsing, discharge cleaning, non-tritium operations and tritium operations are also taken into consideration. A cost estimate is presented

  19. 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.

  20. 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

  1. Thermo-mechanical analysis of ITER first mirrors and its use for the ITER equatorial visible/infrared wide angle viewing system optical design

    International Nuclear Information System (INIS)

    Joanny, M.; Salasca, S.; Dapena, M.; Cantone, B.; Travère, J. M.; Thellier, C.; Fermé, J. J.; Marot, L.; Buravand, O.; Perrollaz, G.; Zeile, C.

    2012-01-01

    ITER first mirrors (FMs), as the first components of most ITER optical diagnostics, will be exposed to high plasma radiation flux and neutron load. To reduce the FMs heating and optical surface deformation induced during ITER operation, the use of relevant materials and cooling system are foreseen. The calculations led on different materials and FMs designs and geometries (100 mm and 200 mm) show that the use of CuCrZr and TZM, and a complex integrated cooling system can limit efficiently the FMs heating and reduce their optical surface deformation under plasma radiation flux and neutron load. These investigations were used to evaluate, for the ITER equatorial port visible/infrared wide angle viewing system, the impact of the FMs properties change during operation on the instrument main optical performances. The results obtained are presented and discussed.

  2. Vacuum system for HIMAC synchrotrons

    International Nuclear Information System (INIS)

    Kanazawa, M.; Sudou, M.; Sato, K.

    1994-01-01

    HIMAC synchrotrons are now under construction, which require vacuum chambers of large aperture and high vacuum of about 10 -9 torr. Wide thin wall vacuum chamber of 0.3 mm thickness reinforced with ribs has been developed as the chamber at dipole magnet. We have just now started to evacuate the lower ring. The obtained average value was about 5x10 -8 torr with turbo-molecular and sputter ion pumps, and 1.1x10 -9 torr after baking. (author)

  3. The upgrade of the DIII-D EC system using 120 GHz ITER gyrotrons

    International Nuclear Information System (INIS)

    Callis, R.W.; Lohr, J.; Gorelov, I.A.; Ponce, D.; Kajiwara, K.; Tooker, J.F.

    2005-01-01

    The planned growth in the EC system on DIII-D over the next few years requires the installation of two depressed collector gyrotrons, a high voltage power supply, two low loss transmission lines, and the required support equipment. This new DIII-D EC equipment could be made identical to the ITER EC system requirements. By building the DIII-D hardware to the ITER specifications, it will allow ITER to gain beneficial prototyping experience on a working tokamak, prior to committing to building the hardware for delivery to ITER

  4. An implicit iterative scheme for solving large systems of linear equations

    International Nuclear Information System (INIS)

    Barry, J.M.; Pollard, J.P.

    1986-12-01

    An implicit iterative scheme for the solution of large systems of linear equations arising from neutron diffusion studies is presented. The method is applied to three-dimensional reactor studies and its performance is compared with alternative iterative approaches

  5. The design and structure of the ultra-high vacuum system of HIRFL-CSR

    International Nuclear Information System (INIS)

    Yang Xiaotian; Zhang Junhui; Zhang Xinjun; Meng Jun; Zhan Wenlong

    2001-01-01

    To minimize the beam loss due to charge exchange of very heavy ions with the residual gas molecules, ultra-high vacuum of 6x10 -9 Pa is required for the HIRFL-CSR facility, which is the lowest pressure in a large vacuum system in China up to now. The total length of the system is about 450 meters and the total inner surface is about 263 square meters. More than 500 standard vacuum components are needed and more than 400 different chambers have to be manufactured. A lot of researches have been down to try to find out the experiences to obtain the required pressure. In this article the following contents are described: the layout of the system; the structure of main vacuum chambers; the treatment method to reduce the outgassing rate of the chamber wall surfaces; vacuum equipment; pressure distribution and the progress of the system

  6. Maintenance implications of critical components in ITER CXRS upper port plug design

    International Nuclear Information System (INIS)

    Koning, Jarich; Jaspers, Roger; Doornink, Jan; Ouwehand, Bernard; Klinkhamer, Friso; Snijders, Bart; Sadakov, Sergey; Heemskerk, Cock

    2009-01-01

    Already in the early phase of a design for ITER, the maintenance aspects should be taken into account, since they might have serious implications. This paper presents the arguments in support of the case for the maintainability of the design, notably if this maintenance is to be performed by advanced remote methods. This structure is compliant to the evolving maintenance strategy of ITER. Initial results of a Failure Mode Effects and Criticality Analysis (FMECA) and a development risk analysis for the ITER upper port plug no. 3, housing the Charge Exchange Recombination Spectroscopy (CXRS) diagnostic, are employed for the definition of the maintenance strategy. The CXRS upper port plug is essentially an optical system which transfers visible light from the plasma into a fiber bundle. The most critical component in this path is the first mirror (M1) whose reflectivity degrades during operation due to deposition and/or erosion dominated effects. Amongst other measures to mitigate these effects, the strategy is to allow for a replacement of this mirror. Therefore it is mounted on a retractable central tube. The main purpose of this tube is to make frequent replacements possible without hindering operation. The maintenance method in terms of time, geometry and spare part policy has a large impact on cost of the system and time usage in the hot cell. Replacement of the tube under vacuum and magnetic field seems infeasible due to the operational risk involved. The preferred solution is to have a spare tube available which is replaced in parallel with other maintenance operations on the vessel, as to avoid any interference in the hot cell with the shutdown scheduling. This avoids having to refurbish a full port plug and also allows for a more frequent replacement of M1, as we can replace the mirror anytime the vacuum vessel is vented, estimated to be once a year.

  7. Vacuum system problems of EBT: a steady-state fusion experiment

    International Nuclear Information System (INIS)

    Livesey, R.L.

    1981-01-01

    Many of the vacuum problems faced by EBT will soon be shared by other plasma devices as high-power microwave systems and long pulse lengths become more common. The solutions used on EBT (such as the raised lip with elastomer seal) are not unique; however, experience has shown that microwave-compatible designs must be carefully thought out. All details of the vacuum must be carefully thought out. All details of the vacuum must be carefully screened in advance to insure that microwaves do not leak into pumps or diagnostics where they can cause major damage. Sputter coating, which even now is noticeably present in most pulsed plasma systems, becomes much worse as systems approach steady state. And finally, radiation degradation of components which is presently a minor problem will become significant on high-power microwave-fed devices, such as EBT-P

  8. Vacuum System and Modeling for the Materials Plasma Exposure Experiment

    International Nuclear Information System (INIS)

    Lumsdaine, Arnold; Meitner, Steve; Graves, Van; Bradley, Craig; Stone, Chris

    2017-01-01

    Understanding the science of plasma-material interactions (PMI) is essential for the future development of fusion facilities. The design of divertors and first walls for the next generation of long-pulse fusion facilities, such as a Fusion Nuclear Science Facility (FNSF) or a DEMO, requires significant PMI research and development. In order to meet this need, a new linear plasma facility, the Materials Plasma Exposure Experiment (MPEX) is proposed, which will produce divertor relevant plasma conditions for these next generation facilities. The device will be capable of handling low activation irradiated samples and be able to remove and replace samples without breaking vacuum. A Target Exchange Chamber (TEC) which can be disconnected from the high field environment in order to perform in-situ diagnostics is planned for the facility as well. The vacuum system for MPEX must be carefully designed in order to meet the requirements of the different heating systems, and to provide conditions at the target similar to those expected in a divertor. An automated coupling-decoupling (“autocoupler”) system is designed to create a high vacuum seal, and will allow the TEC to be disconnected without breaking vacuum in either the TEC or the primary plasma materials interaction chamber. This autocoupler, which can be actuated remotely in the presence of the high magnetic fields, has been designed and prototyped, and shows robustness in a variety of conditions. The vacuum system has been modeled using a simplified finite element analysis, and indicates that the design goals for the pressures in key regions of the facility are achievable.

  9. Summary of beryllium qualification activity for ITER first-wall applications

    International Nuclear Information System (INIS)

    Barabash, V; Eaton, R; Hirai, T; Kupriyanov, I; Nikolaev, G; Wang Zhanhong; Liu Xiang; Roedig, M; Linke, J

    2011-01-01

    Beryllium is considered as an armor material for the ITER first wall. The ITER Final Design Report 2001 identified the reference grades S-65C vacuum hot pressed (VHP) from Brush Wellman and DShG-200 from the Russian Federation. These grades have been selected based on excellent thermal fatigue/shock behavior and the available comprehensive database. Later, Chinese and Russian ITER Parties proposed their new grades: CN-G01 (from China) and TGP-56FW (from Russia). To assess the performance of these new grades, the ITER Organization, Chinese and Russian Parties established a program for the characterization of these materials. A summary of the published data and new results are presented in the paper. It was concluded that the proposed Chinese (CN-G01) and Russian (TGP-56FW) beryllium grades can be accepted. Three grades of beryllium are now available for the armor application for the ITER first wall: S-65, CN-G01 and TGP-56FW.

  10. Summary of beryllium qualification activity for ITER first-wall applications

    Science.gov (United States)

    Barabash, V.; Eaton, R.; Hirai, T.; Kupriyanov, I.; Nikolaev, G.; Wang, Zhanhong; Liu, Xiang; Roedig, M.; Linke, J.

    2011-12-01

    Beryllium is considered as an armor material for the ITER first wall. The ITER Final Design Report 2001 identified the reference grades S-65C vacuum hot pressed (VHP) from Brush Wellman and DShG-200 from the Russian Federation. These grades have been selected based on excellent thermal fatigue/shock behavior and the available comprehensive database. Later, Chinese and Russian ITER Parties proposed their new grades: CN-G01 (from China) and TGP-56FW (from Russia). To assess the performance of these new grades, the ITER Organization, Chinese and Russian Parties established a program for the characterization of these materials. A summary of the published data and new results are presented in the paper. It was concluded that the proposed Chinese (CN-G01) and Russian (TGP-56FW) beryllium grades can be accepted. Three grades of beryllium are now available for the armor application for the ITER first wall: S-65, CN-G01 and TGP-56FW.

  11. Ultra high vacuum system for Isabelle full cell

    International Nuclear Information System (INIS)

    Skelton, R.; Briggs, J.; Chou, T.S.; Foerster, C.; Stattel, P.

    1979-01-01

    A vacuum system consisting of a 40 m long 8.8 cm diameter stainless steel tube, pumped by 7 pumping stations, has been assembled using automatic welding methods. All components have been fired at 950 0 C in a vacuum furnace at a pressure -4 Torr. Each pumping station contains a Ti-sublimator, a 30 liter/s ion pump and an UHV gauge. After assembly, the entire system was baked out at 250 0 C for 24 hours. A pressure -11 Torr was reached after titanium flash. Surface treatment of stainless for 10 -11 Torr operation, bake out and conditioning cycle to read 1 x 10 -11 Torr, and leak checking at low pressures are discussed

  12. Development of core sampling technique for ITER Type B radwaste

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. G.; Hong, K. P.; Oh, W. H.; Park, M. C.; Jung, S. H.; Ahn, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Type B radwaste (intermediate level and long lived radioactive waste) imported from ITER vacuum vessel are to be treated and stored in basement of hot cell building. The Type B radwaste treatment process is composed of buffer storage, cutting, sampling/tritium measurement, tritium removal, characterization, pre-packaging, inspection/decontamination, and storage etc. The cut slices of Type B radwaste components generated from cutting process undergo sampling process before and after tritium removal process. The purpose of sampling is to obtain small pieces of samples in order to investigate the tritium content and concentration of Type B radwaste. Core sampling, which is the candidates of sampling technique to be applied to ITER hot cell, is available for not thick (less than 50 mm) metal without use of coolant. Experimented materials were SS316L and CuCrZr in order to simulate ITER Type B radwaste. In core sampling, substantial secondary wastes from cutting chips will be produced unavoidably. Thus, core sampling machine will have to be equipped with disposal system such as suction equipment. Core sampling is considered an unfavorable method for tool wear compared to conventional drilling.

  13. The ITER Plasma Control System Simulation Platform

    International Nuclear Information System (INIS)

    Walker, M.L.; Ambrosino, G.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W.; Welander, A.S.; Winter, A.

    2015-01-01

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  14. The ITER Plasma Control System Simulation Platform

    Energy Technology Data Exchange (ETDEWEB)

    Walker, M.L., E-mail: walker@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Ambrosino, G.; De Tommasi, G. [CREATE/Università di Napoli Federico II, Napoli (Italy); Humphreys, D.A. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Mattei, M. [CREATE/Seconda Università di Napoli, Napoli (Italy); Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Welander, A.S. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Winter, A. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

    2015-10-15

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  15. Natural Preconditioning and Iterative Methods for Saddle Point Systems

    KAUST Repository

    Pestana, Jennifer

    2015-01-01

    © 2015 Society for Industrial and Applied Mathematics. The solution of quadratic or locally quadratic extremum problems subject to linear(ized) constraints gives rise to linear systems in saddle point form. This is true whether in the continuous or the discrete setting, so saddle point systems arising from the discretization of partial differential equation problems, such as those describing electromagnetic problems or incompressible flow, lead to equations with this structure, as do, for example, interior point methods and the sequential quadratic programming approach to nonlinear optimization. This survey concerns iterative solution methods for these problems and, in particular, shows how the problem formulation leads to natural preconditioners which guarantee a fast rate of convergence of the relevant iterative methods. These preconditioners are related to the original extremum problem and their effectiveness - in terms of rapidity of convergence - is established here via a proof of general bounds on the eigenvalues of the preconditioned saddle point matrix on which iteration convergence depends.

  16. Accuracy analysis of hybrid parallel robot for the assembling of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Wang Yongbo [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland); The State Key Laboratory of Mechanical Transmission, Chongqing University (China); Pessi, Pekka [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland); Wu Huapeng [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland)], E-mail: huapeng@lut.fi; Handroos, Heikki [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland)

    2009-06-15

    This paper presents a novel mobile parallel robot, which is able to carry welding and machining processes from inside the international thermonuclear experimental reactor (ITER) vacuum vessel (VV). The kinematics design of the robot has been optimized for ITER access. To improve the accuracy of the parallel robot, the errors caused by the stiffness and manufacture process have to be compensated or limited to a minimum value. In this paper kinematics errors and stiffness modeling are given. The simulation results are presented.

  17. Accuracy analysis of hybrid parallel robot for the assembling of ITER

    International Nuclear Information System (INIS)

    Wang Yongbo; Pessi, Pekka; Wu Huapeng; Handroos, Heikki

    2009-01-01

    This paper presents a novel mobile parallel robot, which is able to carry welding and machining processes from inside the international thermonuclear experimental reactor (ITER) vacuum vessel (VV). The kinematics design of the robot has been optimized for ITER access. To improve the accuracy of the parallel robot, the errors caused by the stiffness and manufacture process have to be compensated or limited to a minimum value. In this paper kinematics errors and stiffness modeling are given. The simulation results are presented.

  18. 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)

  19. Challenges of ITER diagnostic electrical services

    Energy Technology Data Exchange (ETDEWEB)

    Encheva, A., E-mail: anna.encheva@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Omran, H. [Oxford Technologies Ltd, 7 Nuffield Way, Abingdon OX14 1RL (United Kingdom); Pérez-Lasala, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, n° 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain); Alekseev, A. [Efremov Institute, Metallostroy, Doroga na Metallostroy, 3 bld., Saint-Petersburg 196641 (Russian Federation); Arshad, S. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, n° 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain); Bede, O. [Oxford Technologies Ltd, 7 Nuffield Way, Abingdon OX14 1RL (United Kingdom); Bender, S. [Efremov Institute, Metallostroy, Doroga na Metallostroy, 3 bld., Saint-Petersburg 196641 (Russian Federation); Bertalot, L.; Direz, M.-F.; Drevon, J.-M.; Jakhar, S.; Kaschuk, Y.; Komarov, V.; Lebarbier, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Lucca, F. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Macklin, B.; Maquet, P. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Marin, A. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Martin, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Mills, S. [Oxford Technologies Ltd, 7 Nuffield Way, Abingdon OX14 1RL (United Kingdom); and others

    2013-10-15

    Highlights: • A brief description of all major components part of diagnostic electrical services has been given. • The integration challenges have been presented. • Design assumptions and requirements for the components have been described. • The design of the conduit/loom and the relevant analysis has been highlighted. -- Abstract: Diagnostic electrical services provide the electrical infrastructure to serve diagnostic components installed on the ITER tokamak. This infrastructure is composed of cables, connectors, cable tails, looms, conduits and feedthroughs. The diagnostic services offer as well a shelter for various instrumentations – vacuum vessel (VV), blanket and divertor. The diagnostic sensors are located on the inner and outer VV wall, on blanket shield modules, divertor cassettes and in port plugs. They require electrical cabling to extract the measurement and, in some cases, to supply electrical power to the sensors. These cables run from the sensors to feedthroughs on the VV and the port interspace or cryostat. The design and integration of all components that are part of diagnostic electrical services is an important engineering activity that is being challenged by the multiple requirements and constraints which have to be satisfied while at the same time delivering the required diagnostic performance. The positioning of the components must correlate not only with their functional specifications but also with the design of the major ITER components. This is a particular challenge because not all systems have reached the same level of design maturity. This paper outlines the engineering challenges of ITER diagnostics electrical services. The environmental conditions inside the VV will have an important impact. Leading risks to these components include poor electrical contact at connectors, the effects of exposure to nuclear irradiation, such as material transmutation, heating, and generation of spurious electrical signals etc., failure due to

  20. Investigation and analysis on ITER in-vessel coils’ raw-materials

    International Nuclear Information System (INIS)

    Jin, Huan; Wu, Yu; Long, Feng; Yu, Min; Han, Qiyang; Liu, Huajun

    2013-01-01

    Highlights: • The R and D works for the ITER in-vessel coils (IVC) are now being conducted in Institute of Plasma Physics, and the analysis work are being done by Princeton Plasma Physics Laboratory. • There is little published paper about the raw materials for ITER IVC coils. • This manuscript points out the progress of the selected materials for ITER IVC coils. -- Abstract: The ITER in-vessel coils (IVCs) consist of 27 coils edge localized modes (ELM) and 2 coils vertical stabilization (VS) which are all mounted on the vacuum vessel wall behind the shield modules. The IVCs design and manufacturing work is being conducted in between Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and Princeton Plasma Physics Laboratory (PPPL). Because the position of ELM and VS coils is close and face to the plasma, the IVCs must undergo a severe environment, such as the high dose of radiation and high operation temperature, thus the conventional electrical insulation materials cannot be used. And the technology of “Stainless Steel Jacketed Mineral Insulated Conductor” (SSMIC) is deemed as the best choice to provide the necessary radiation resistance and compatibility strength in ITER's vacuum vessel. While mineral insulated conductor technology is not new, and is similar to the mineral insulated cable used in industrial. Some difficulties still need to be solved, such as searching for the proper raw-materials to make sure that the conductor have the properties of high current carrying capability, the necessary radiation resistance, the proper strength, at the same time, it must be come true in manufacture technology. This paper described the analysis of the materials for VS and ELM coil conductor

  1. Thermo-mechanical design methodology for ITER cryo-distribution cold boxes

    International Nuclear Information System (INIS)

    Shukla, Vinit; Patel, Pratik; Vaghela, Hiten; Das, Jotirmoy; Shah, Nitin; Bhattacharya, Ritendra; Sarkar, Biswanath; Chang, Hyun-sik

    2015-01-01

    The ITER cryo-distribution system is in charge of the proper distribution of the cryogen at required mass flow rate, pressure and temperature level to the users namely; the superconducting magnets and cryopumps. The cryo-distribution also acts as a thermal buffer in order to run the cryo-plant as much as possible at a steady state condition. A typical cryo-distribution cold box is equipped with mainly liquid helium bath with heat exchangers, cryogenic valves, cold circulating pump and cold compressor. During the intended operation life of ITER, several loads on the cryo-distribution system are envisaged, these are, gravity/assembly loads, nominal pressure/temperature, test pressure/temperature, purge pressure, thermo-mechanical loads due to break of insulation vacuum, transport acceleration and seismic loads. Single loads or combinations of them can act on the cryo-distribution system and its components; therefore, it is very important to analyze the behavior of the system and components under the influence of these loads or combinations. Possible load combinations for the cryo-distribution system will be analyzed and will lead to the basis of the design. This paper will focus on the understanding of the nature of the loads and their combinations for the ITER cryo-distribution as well as their impacts on the design. A representative model of a cold box is considered on which the load combinations have been applied in order to understand their impacts on the design of the cryo-distribution. Also the worst-impact loads or their combination which drive the design of cryo-distribution cold boxes will be derived. (author)

  2. Approximate convex hull of affine iterated function system attractors

    International Nuclear Information System (INIS)

    Mishkinis, Anton; Gentil, Christian; Lanquetin, Sandrine; Sokolov, Dmitry

    2012-01-01

    Highlights: ► We present an iterative algorithm to approximate affine IFS attractor convex hull. ► Elimination of the interior points significantly reduces the complexity. ► To optimize calculations, we merge the convex hull images at each iteration. ► Approximation by ellipses increases speed of convergence to the exact convex hull. ► We present a method of the output convex hull simplification. - Abstract: In this paper, we present an algorithm to construct an approximate convex hull of the attractors of an affine iterated function system (IFS). We construct a sequence of convex hull approximations for any required precision using the self-similarity property of the attractor in order to optimize calculations. Due to the affine properties of IFS transformations, the number of points considered in the construction is reduced. The time complexity of our algorithm is a linear function of the number of iterations and the number of points in the output approximate convex hull. The number of iterations and the execution time increases logarithmically with increasing accuracy. In addition, we introduce a method to simplify the approximate convex hull without loss of accuracy.

  3. Development of a Remote Handling Robot for the Maintenance of an ITER-Like D-Shaped Vessel

    Directory of Open Access Journals (Sweden)

    Peihua Chen

    2014-01-01

    Full Text Available Robotic operation is one of the major challenges in the remote maintenance of ITER vacuum vessel (VV and future fusion reactors as inner operations of Tokamak have to be done by robots due to the internal adverse conditions. This paper introduces a novel remote handling robot (RHR for the maintenance of ITER-like D-shaped vessel. The modular designed RHR, which is an important part of the remote handling system for ITER, consists of three parts: an omnidirectional transfer vehicle (OTV, a planar articulated arm (PAA, and an articulated teleoperated manipulator (ATM. The task of RHR is to carry processing tools, such as the viewing system, leakage detector, and electric screwdriver, to inspect and maintain the components installed inside the D-shaped vessel. The kinematics of the OTV, as well as the kinematic analyses of the PAA and ATM, is studied in this paper. Because of its special length and heavy payload, the dynamics of the PAA is also investigated through a dynamic simulation system based on robot technology middleware (RTM. The results of the path planning, workspace simulations, and dynamic simulation indicate that the RHR has good mobility together with satisfying kinematic and dynamic performances and can well accomplish its maintenance tasks in the ITER-like D-shaped vessel.

  4. Image based rendering of iterated function systems

    NARCIS (Netherlands)

    Wijk, van J.J.; Saupe, D.

    2004-01-01

    A fast method to generate fractal imagery is presented. Iterated function systems (IFS) are based on repeatedly copying transformed images. We show that this can be directly translated into standard graphics operations: Each image is generated by texture mapping and blending copies of the previous

  5. Development of design Criteria for ITER In-vessel Components

    International Nuclear Information System (INIS)

    Sannazzaro, G.; Barabash, V.; Kang, S.C.; Fernandez, E.; Kalinin, G.; Obushev, A.; Martínez, V.J.; Vázquez, I.; Fernández, F.; Guirao, J.

    2013-01-01

    Absrtract: The components located inside the ITER vacuum chamber (in-vessel components – IC), due to their specific nature and the environments they are exposed to (neutron radiation, high heat fluxes, electromagnetic forces, etc.), have specific design criteria which are, in this paper, referred as Structural Design Criteria for In-vessel Components (SDC-IC). The development of these criteria started in the very early phase of the ITER design and followed closely the criteria of the RCC-MR code. Specific rules to include the effect of neutron irradiation were implemented. In 2008 the need of an update of the SDC-IC was identified to add missing specifications, to implement improvements, to modernise rules including recent evolutions in international codes and regulations (i.e. PED). Collaboration was set up between ITER Organization (IO), European (EUDA) and Russian Federation (RFDA) Domestic Agencies to generate a new version of SDC-IC. A Peer Review Group (PRG) composed by members of the ITER Organization and all ITER Domestic Agencies and code experts was set-up to review the proposed modifications, to provide comments, contributions and recommendations

  6. LSODKR, Stiff Ordinary Differential Equations (ODE) System Solver with Krylov Iteration with Root-finding

    International Nuclear Information System (INIS)

    Hindmarsh, A.C.; Petzold, L.R.

    2005-01-01

    1 - Description of program or function: LSODKR is a new initial value ODE solver for stiff and non-stiff systems. It is a variant of the LSODPK and LSODE solvers, intended mainly for large stiff systems. The main differences between LSODKR and LSODE are the following: a) for stiff systems, LSODKR uses a corrector iteration composed of Newton iteration and one of four preconditioned Krylov subspace iteration methods. The user must supply routines for the preconditioning operations, b) within the corrector iteration, LSODKR does automatic switching between functional (fix point) iteration and modified Newton iteration, The nonlinear iteration method-switching differs from the method-switching in LSODA and LSODAR, but provides similar savings by using the cheaper method in the non-stiff regions of the problem. c) LSODKR includes the ability to find roots of given functions of the solution during the integration. d) LSODKR also improves on the Krylov methods in LSODPK by offering the option to save and reuse the approximate Jacobian data underlying the pre-conditioner. The LSODKR source is commented extensively to facilitate modification. Both a single-precision version and a double-precision version are available. 2 - Methods: It is assumed that the ODEs are given explicitly, so that the system can be written in the form dy/dt = f(t,y), where y is the vector of dependent variables, and t is the independent variable. Integration is by Adams or BDF (Backward Differentiation Formula) methods, at user option. Corrector iteration is by Newton or fix point iteration, determined dynamically. Linear system solution is by a preconditioned Krylov iteration, selected by user from Incomplete Orthogonalization Method, Generalized Minimum Residual Method, and two variants of Preconditioned Conjugate Gradient Method. Preconditioning is to be supplied by the user

  7. 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)

  8. Data archiving system implementation in ITER's CODAC Core System

    Energy Technology Data Exchange (ETDEWEB)

    Castro, R., E-mail: rodrigo.castro@visite.es [CIEMAT Fusion Program, Avda. Complutense 40, Madrid (Spain); Abadie, L. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France); Makushok, Y. [Sgenia, C/Chile, 4 Edificio II, Las Rozas, Madrid (Spain); Ruiz, M.; Sanz, D. [Instrumentation and Applied Acoustic Research Group, Technical University of Madrid, Madrid (Spain); Vega, J. [CIEMAT Fusion Program, Avda. Complutense 40, Madrid (Spain); Faig, J. [INDRA Sistemas, S.A. Unid. de Sistemas de Control, Dirección de Tecnología Energética, Madrid (Spain); Román-Pérez, G. [Sgenia, C/Chile, 4 Edificio II, Las Rozas, Madrid (Spain); Simrock, S.; Makijarvi, P. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

    2015-10-15

    Highlights: • Implementation of ITER's data archiving solution. • Integration of the solution into CODAC Core System. • Data archiving structure. • High efficient data transmission into fast plant system controllers. • Fast control and data acquisition in Linux. - Abstract: The aim of this work is to present the implementation of data archiving in ITER's CODAC Core System software. This first approach provides a client side API and server side software allowing the creation of a simplified version of ITERDB data archiving software, and implements all required elements to complete data archiving flow from data acquisition until its persistent storage technology. The client side includes all necessary components that run on devices that acquire or produce data, distributing and streaming to configure remote archiving servers. The server side comprises an archiving service that stores into HDF5 files all received data. The archiving solution aims at storing data coming for the data acquisition system, the conventional control and also processed/simulated data.

  9. ITER Organization - 2012 Annual Report, 2012 Financial Statements

    International Nuclear Information System (INIS)

    2013-01-01

    In its first part, this report gives an overview of the main activities and events regarding the ITER organization, the ITER project baseline, the construction of seismic foundations, the licensing decree in France, the procurement arrangements, the manufacturing of the ITER vacuum vessel, the research and development for prototype development, the management of member contributions in France, the creation of new positions as far as staffing is concerned. The second part presents the various highlights for the year and by department: Office of the Director-General, Legal Affairs, International Audit, ITER Council Secretariat, Bureau of International Cooperation, Department for ITER Project (Directorate for Central Integration and Engineering, Directorate for Tokamak, Directorate for CODAC, Heating and Diagnostics, Directorate for Buildings and Site Infrastructure, Directorate for Central Engineering and Plant, Directorate for Plasma Operation), Department for Safety Quality and Security, Department for Administration). The next parts contain tables and charts which present staffing and financial data, presentations of procurement highlights and data for domestic agencies (China, Europe, India, Japan, Korea, Russia, USA) in terms of R and D and manufacturing, of contracts. The last part presents and comments the financial statements for 2012

  10. Generic Diagnostic Port Integration for the Equatorial Port Plug of ITER

    International Nuclear Information System (INIS)

    Doceul, L.; Chappuis, Ph.; Portafaix, Ch.; Guillaume, T.; Bruyere, Ch.; Walker, Ch.; Ingesson, Ch.; Ciattaglia, E.; Salasca, S.; Eric, T.

    2006-01-01

    ITER requires an extensive set of diagnostic systems to provide several key functions such as protection of the device, input to plasma control systems and evaluation of the plasma performance. Most of these diagnostics system are to be integrated in port plugs, which are water cooled stainless steel structures (approximately: 50 t, 2 m x 2 m x 4 m) inserted into the vacuum-vessel ports. The port plug must perform basic functions such as providing neutron and gamma shielding, supporting the first wall armour and shielding blanket material, closing the vacuum vessel ports, supporting the diagnostic equipment (within the primary vacuum, on the primary vacuum boundary and in the port interspace). CEA (Commissariat l'Energie Atomique) has contributed to the engineering activities on the port plugs and has more particularly focused on the design and diagnostic integration in the representative equatorial port plug EQ01. The specific CEA contributions were to perform the general engineering, structural and thermal analysis. These detailed analysis have highlighted some design issues which were worked out through different solutions. This paper will contain the description of the engineering activities performed such as: - The conceptual design of the EQ01 and the associated diagnostics, such as the visible and infrared optical diagnostic, - The static mechanical calculations, taking into account the electromagnetic loads occurring during fast transient plasma events, - The dynamic calculation constituted of modal and transient analysis under the same electromagnetic loads to estimate the dynamic amplification factor due to the resonance phenomenon, - The thermal assessment under the neutronic load of the water-cooled stainless steel structure, - The seismic response of the port plug inside the vacuum vessel, taking into account the ground spectra and soil conditions in the Cadarache site. (author)

  11. Upgrading of TARN-II vacuum system

    International Nuclear Information System (INIS)

    Chida, K.; Arakaki, Y.; Yoshizawa, M.; Tomizawa, M.; Tanabe, T.; Katayama, I.

    1994-01-01

    Ion pumps and titanium getter pumps have been increased nearly twice in the TARN-II. The pumping speed per unit length is now improved up to 2/3 times that of TARN-I. An average vacuum pressure of 10 -11 Torr order has been achieved at beam time. Performance of the system after the upgrading is reported. (author)

  12. APS storage ring vacuum system performance

    International Nuclear Information System (INIS)

    Noonan, J.R.; Gagliano, J.; Goeppner, G.A.

    1997-01-01

    The Advanced Photon Source (APS) storage ring was designed to operated with 7-GeV, 100-mA positron beam with lifetimes > 20 hours. The lifetime is limited by residual gas scattering and Touschek scattering at this time. Photon-stimulated desorption and microwave power in the rf cavities are the main gas loads. Comparison of actual system gas loads and design calculations will be given. In addition, several special features of the storage ring vacuum system will be presented

  13. Physics of the conceptual design of the ITER plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Snipes, J.A., E-mail: Joseph.Snipes@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bremond, S. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Campbell, D.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Casper, T. [1166 Bordeaux St, Pleasanton, CA 94566 (United States); Douai, D. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Gribov, Y. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Humphreys, D. [General Atomics, San Diego, CA 92186 (United States); Lister, J. [Association EURATOM-Confédération Suisse, Ecole Polytechnique Fédérale de Lausanne (EPFL), CRPP, Lausanne CH-1015 (Switzerland); Loarte, A.; Pitts, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sugihara, M., E-mail: Sugihara_ma@yahoo.co.jp [Japan (Japan); Winter, A.; Zabeo, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

    2014-05-15

    Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

  14. Physics of the conceptual design of the ITER plasma control system

    International Nuclear Information System (INIS)

    Snipes, J.A.; Bremond, S.; Campbell, D.J.; Casper, T.; Douai, D.; Gribov, Y.; Humphreys, D.; Lister, J.; Loarte, A.; Pitts, R.; Sugihara, M.; Winter, A.; Zabeo, L.

    2014-01-01

    Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

  15. Vacuum mechatronics first international workshop

    Energy Technology Data Exchange (ETDEWEB)

    Belinski, S.E.; Shirazi, M.; Hackwood, S.; Beni, G. (eds.) (California Univ., Santa Barbara, CA (USA))

    1989-01-01

    This report contains papers on the following topics: proposed epitaxial thin film growth in the ultra-vacuum of space; particle monitoring and control in vacuum processing equipment; electrostatic dust collector for use in vacuum systems; materials evaluation of an electrically noisy vacuum slip ring assembly; an overview of lubrication and associated materials for vacuum service; the usage of lubricants in a vacuum environment; guidelines and practical applications for lubrication in vacuum; recent development in leak detector and calibrator designs; the durability of ballscrews for ultrahigh vacuum; vacuum-compatible robot for self-contained manufacturing systems; the design, fabrication, and assembly of an advanced vacuum robotics system for space payload calibration; design criteria for mechanisms used in space; and concepts and requirements for semiconductor multiprocess integration in vacuum. These papers are indexed separately elsewhere.

  16. 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

  17. Fabrication of a full-size mock-up for inboard 10o section of ITER vacuum vessel thermal shield

    International Nuclear Information System (INIS)

    Chung, W.; Nam, K.; Noh, C.H.; Kang, D.K.; Kang, S.M.; Oh, Y.G.; Choi, S.W.; Kang, S.H.; Utin, Y.; Ioki, K.; Her, N.; Yu, J.

    2011-01-01

    A full-scale mock-up of VVTS inboard section was made in order to validate its manufacturing processes before manufacturing the vacuum vessel thermal shield (VVTS) for ITER tokamak. VVTS inboard 10 o section consists of 20 mm shells on which cooling tubes are welded and flange joints that connect adjacent thermal shield sectors. The whole VVTS inboard is divided into two by bisectional flange joint located at the center. All the manufacturing processes except silver coating were tested and verified in the fabrication of mock-up. For the forming and the welding, pre-qualification tests were conducted to find proper process conditions. Shell thickness change was measured after bending, forming and buffing processes. Shell distortion was adjusted after the welding. Welding was validated by non-destructive examination. Bisectional flange joint was successfully assembled by inserting pins and tightening with bolt/nut. Bolt hole margin of 2 mm for sector flange was revealed to be sufficient by successful sector assembly of upper and lower parts of mock-up. Handling jig was found to be essential because the inboard section was flexible. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner.

  18. Iterative group splitting algorithm for opportunistic scheduling systems

    KAUST Repository

    Nam, Haewoon; Alouini, Mohamed-Slim

    2014-01-01

    An efficient feedback algorithm for opportunistic scheduling systems based on iterative group splitting is proposed in this paper. Similar to the opportunistic splitting algorithm, the proposed algorithm adjusts (or lowers) the feedback threshold

  19. Requirements and guidelines for NSLS experimental beam line vacuum systems: Revision A

    International Nuclear Information System (INIS)

    Foerster, C.; Halama, H.; Thomlinson, W.

    1986-10-01

    Requirements are provided for NSLS beam line front ends and vacuum interlocks. Guidelines are provided for UHV beam line vacuum systems, including materials, vacuum hardware (pumps, valves, and flanges), acoustic delay lines and beam line fast valves, instrumentation, fabrication and testing, and the NSLS cleaning facility. Also discussed are the design review for experimenters' equipment that would be connected to the NSLS and acceptance tests for any beam line to be connected with the ring vacuum. Also appended are a description of the acoustic delay line as well as the NSLS vacuum standards and NSLS procedures

  20. Availability analysis of the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

    2015-01-01

    The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

  1. The Control System of CERN Accelerators Vacuum (Current Status and Recent Improvements)

    CERN Document Server

    Gomes, P; Blanchard, S; Boccioli, M; Girardot, G; Vestergard, H; Kopylov, L; Mikheev, M

    2011-01-01

    The vacuum control system of most of the CERN accelerators is based on Siemens PLCs and on PVSS SCADA. After the transition from the LHC commissioning phase to its regular operation, there has been a number of additions and improvements to the vacuum control system. They were driven by new technical requirements and by feedback from the accelerator operators and vacuum specialists. New control functions have been implemented in the PLCs; new tools have been developed for the SCADA, while its ergonomics and navigation have been enhanced.

  2. EAST-AIA deployment under vacuum: Calibration of laser diagnostic system using computer vision

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yang, E-mail: yangyang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, 350 Shushanhu Rd, Hefei, Anhui (China); Song, Yuntao; Cheng, Yong; Feng, Hansheng; Wu, Zhenwei; Li, Yingying; Sun, Yongjun; Zheng, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, 350 Shushanhu Rd, Hefei, Anhui (China); Bruno, Vincent; Eric, Villedieu [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France)

    2016-11-15

    Highlights: • The first deployment of the EAST articulated inspection arm robot under vacuum is presented. • A computer vision based approach to measure the laser spot displacement is proposed. • An experiment on the real EAST tokamak is performed to validate the proposed measure approach, and the results shows that the measurement accuracy satisfies the requirement. - Abstract: For the operation of EAST tokamak, it is crucial to ensure that all the diagnostic systems are in the good condition in order to reflect the plasma status properly. However, most of the diagnostic systems are mounted inside the tokamak vacuum vessel, which makes them extremely difficult to maintain under high vacuum condition during the tokamak operation. Thanks to a system called EAST articulated inspection arm robot (EAST-AIA), the examination of these in-vessel diagnostic systems can be performed by an embedded camera carried by the robot. In this paper, a computer vision algorithm has been developed to calibrate a laser diagnostic system with the help of a monocular camera at the robot end. In order to estimate the displacement of the laser diagnostic system with respect to the vacuum vessel, several visual markers were attached to the inner wall. This experiment was conducted both on the EAST vacuum vessel mock-up and the real EAST tokamak under vacuum condition. As a result, the accuracy of the displacement measurement was within 3 mm under the current camera resolution, which satisfied the laser diagnostic system calibration.

  3. ITER cooling system

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-11-01

    The present specification of the ITER cooling system does not permit its operation with water above 150 C. However, the first wall needs to be heated to higher temperatures during conditioning at 250 C and bake-out at 350 C. In order to use the cooling water for these operations the cooling system would have to operate during conditioning at 37 Bar and during bake-out at 164 Bar. This is undesirable from the safety analysis point of view, and alternative heating methods are to be found. This review suggests that superheated steam or gas heating can be used for both baking and conditioning. The blanket design must consider the use of dual heat transfer media, allowing for change from one to another in both directions. Transfer from water to gas or steam is the most intricate and risky part of the entire heating process. Superheated steam conditioning appears unfavorable. The use of inert gas is recommended, although alternative heating fluids such as organic coolant should be investigated

  4. Development of a Modified Vacuum Cleaner for Lunar Surface Systems

    Science.gov (United States)

    Toon, Katherine P.; Lee, Steve A.; Edgerly, Rachel D.

    2010-01-01

    The National Aeronautics and Space Administration (NASA) mission to expand space exploration will return humans to the Moon with the goal of maintaining a long-term presence. One challenge that NASA will face returning to the Moon is managing the lunar regolith found on the Moon's surface, which will collect on extravehicular activity (EVA) suits and other equipment. Based on the Apollo experience, the issues astronauts encountered with lunar regolith included eye/lung irritation, and various hardware failures (seals, screw threads, electrical connectors and fabric contamination), which were all related to inadequate lunar regolith mitigation. A vacuum cleaner capable of detaching, transferring, and efficiently capturing lunar regolith has been proposed as a method to mitigate the lunar regolith problem in the habitable environment on lunar surface. In order to develop this vacuum, a modified "off-the-shelf' vacuum cleaner will be used to determine detachment efficiency, vacuum requirements, and optimal cleaning techniques to ensure efficient dust removal in habitable lunar surfaces, EVA spacesuits, and air exchange volume. During the initial development of the Lunar Surface System vacuum cleaner, systematic testing was performed with varying flow rates on multiple surfaces (fabrics and metallics), atmospheric (14.7 psia) and reduced pressures (10.2 and 8.3 psia), different vacuum tool attachments, and several vacuum cleaning techniques in order to determine the performance requirements for the vacuum cleaner. The data recorded during testing was evaluated by calculating particulate removal, relative to the retained simulant on the tested surface. In addition, optical microscopy was used to determine particle size distribution retained on the surface. The scope of this paper is to explain the initial phase of vacuum cleaner development, including historical Apollo mission data, current state-of-the-art vacuum cleaner technology, and vacuum cleaner testing that has

  5. Review of Current Nuclear Vacuum System Technologies

    International Nuclear Information System (INIS)

    Carroll, M.; McCracken, J.; Shope, T.

    2003-01-01

    Nearly all industrial operations generate unwanted dust, particulate matter, and/or liquid wastes. Waste dust and particulates can be readily tracked to other work locations, and airborne particulates can be spread through ventilation systems to all locations within a building, and even vented outside the building - a serious concern for processes involving hazardous, radioactive, or nuclear materials. Several varieties of vacuum systems have been proposed and/or are commercially available for clean up of both solid and liquid hazardous and nuclear materials. A review of current technologies highlights both the advantages and disadvantages of the various systems, and demonstrates the need for a system designed to address issues specific to hazardous and nuclear material cleanup. A review of previous and current hazardous/nuclear material cleanup technologies is presented. From simple conventional vacuums modified for use in industrial operations, to systems specifically engineered for such purposes, the advantages and disadvantages are examined in light of the following criteria: minimal worker exposure; minimal secondary waste generation;reduced equipment maintenance and consumable parts; simplicity of design, yet fully compatible with all waste types; and ease of use. The work effort reviews past, existing and proposed technologies in light of such considerations. Accomplishments of selected systems are presented, including identified areas where technological improvements could be suggested

  6. 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)

  7. Nuclear Analysis of an ITER Blanket Module

    Science.gov (United States)

    Chiovaro, P.; Di Maio, P. A.; Parrinello, V.

    2013-08-01

    ITER blanket system is the reactor's plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.

  8. REQUIREMENTS AND GUIDELINES FOR NSLS EXPERIMENTAL BEAM LINE VACUUM SYSTEMS-REVISION B

    International Nuclear Information System (INIS)

    FOERSTER, C.

    1999-01-01

    Typical beam lines are comprised of an assembly of vacuum valves and shutters referred to as a ''front end'', optical elements to monochromatize, focus and split the photon beam, and an experimental area where a target sample is placed into the photon beam and data from the interaction is detected and recorded. Windows are used to separate sections of beam lines that are not compatible with storage ring ultra high vacuum. Some experimental beam lines share a common vacuum with storage rings. Sections of beam lines are only allowed to vent up to atmospheric pressure using pure nitrogen gas after a vacuum barrier is established to protect ring vacuum. The front end may only be bled up when there is no current in the machine. This is especially true on the VUV storage ring where for most experiments, windows are not used. For the shorter wavelength, more energetic photons of the x-ray ring, beryllium windows are used at various beam line locations so that the monochromator, mirror box or sample chamber may be used in a helium atmosphere or rough vacuum. The window separates ring vacuum from the environment of the downstream beam line components. The stored beam lifetime in the storage rings and the maintenance of desirable reflection properties of optical surfaces depend upon hydrocarbon-free, ultra-high vacuum systems. Storage ring vacuum systems will operate at pressures of ∼ 1 x 10 -10 Torr without beam and ∼ 1 x 10 -9 Torr with beam. Systems are free of hydrocarbons in the sense that no pumps, valves, etc. containing organics are used. Components are all-metal, chemically cleaned and bakeable. To the extent that beam lines share a common vacuum with the storage ring, the same criteria will hold for beam line components. The design philosophy for NSLS beam lines is to use all-metal, hydrocarbon-free front end components and recommend that experimenters use this approach for common vacuum hardware downstream of front ends. O-ring-sealed valves, if used, are not

  9. ITER in-vessel system design and performance

    Science.gov (United States)

    Parker, R. R.

    2000-03-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events.

  10. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    2000-01-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events. (author)

  11. PTC-6 vacuum system: WallWalker{trademark} and Blastrac{reg_sign} shot blast cleaning system

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    The LTC Americas, Inc. wall decontamination technology consisted of two pneumatic hand-held tools: (1) a roto-peen scaler that used star cutters and (2) a 3-piston hammer with reciprocating bits. The hand-held tools were used in conjunction with the LTC PTC-6 vacuum system which captured dust and debris as the wall decontamination took place. Recommendations for improved worker safety and health during use of the PTC-6 vacuum system with hand-held tools include: (1) keeping all hoses and lines as orderly as possible in compliance with good housekeeping requirements; (2) ergonomic training to include techniques in lifting, bending, stooping, twisting, etc.; (3) use of a clamping system to hold hoses to the vacuum system; (4) a safety line on the air line connections; (5) use of a mechanical lifting system for waste drum removal; and (6) the use of ergonomically designed tools.

  12. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H

    2001-11-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

  13. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

    2001-01-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

  14. Enhanced configuration of a water detritiation system; impact on ITER Isotope Separation System based cryogenic distillation

    Energy Technology Data Exchange (ETDEWEB)

    Cristescu, Ion, E-mail: ion.cristescu@kit.edu

    2016-11-01

    Highlights: • An enhanced configuration of ITER WDS has been developed. • The proposed configuration allows minimization of hazards due to the reduction of tritium inventory. • The load on the tritium recovery system (ITER ISS) is minimized with benefits on mitigation of the explosion hazards. - Abstract: Tritiated water is generated in the ITER systems by various sources and may contain deuterium and tritium at various concentrations. The reference process for the ITER Water Detritiation System is based on Combined Electrolysis Catalytic Exchange (CECE) configuration. During long time operation of the CECE process, the accumulation of deuterium in the electrolysis unit and consequently along the Liquid Phase Catalytic Exchange (LPCE) column is unavoidable with consequences on the overall detritiation factor of the system. Beside the deuterium issue in the process, the large amount of the tritiated water with tritium activity up to 500 Ci/kg in the electrolysis cells is a concern from the safety aspect of the plant. The enhanced configuration of a system for processing tritiated water allows mitigation of the effects due to deuterium accumulation and also reduction of tritium inventory within the electrolysis system. In addition the benefits concerning to the interface between the water detritiation system and tritium recovery based cryogenic distillation are also presented.

  15. High heat flux (HHF) elements for negative ion systems on ITER

    International Nuclear Information System (INIS)

    Milnes, J.; Chuilon, B.; Xue, Y.; Martin, D.; Waldon, C.

    2007-01-01

    Negative Ion Neutral Beam systems on ITER will require actively cooled scrapers and dumps to process and shape the beam before injection into the tokamak. The scale of the systems is much larger than any presently operating, bringing challenges for designers in terms of available sub cooling, total pressure drop, deflection and mandatory remote maintenance. High heat fluxes (∼15-20 MW/m 2 ), pulse lengths in excess of 3000 s and high number of cycles pose new challenges in terms of stress and fatigue life. The designs outlined in the Design Description Document for the ITER Neutral Beam System [N53 DDD 29 01-07-03 R 0.1. ITER Design Description Document, DDD 5.3, Neutral Beam H and CD system (including Appendices).], based on swirl tubes, have been reviewed as part of the design process and recommendations made. Additionally, alternative designs have been proposed based on the Hypervapotron high heat flux elements with modified geometry and drawing upon a vast background knowledge of large scale equipment procurement and integration. A full thermo-mechanical analysis of all HHF components has also been undertaken based on ITER design criteria and the limited material data available. The advantages and disadvantages of all designs are presented and recommendations for improvements discussed

  16. The control system of CERN accelerators vacuum (current status and recent improvements)

    International Nuclear Information System (INIS)

    Gomes, P.; Antoniotti, F.; Blanchard, S.; Boccioli, M.; Girardot, G.; Vestergard, H.; Kopylov, L.; Mikheev, M.

    2012-01-01

    The vacuum control system of most of the CERN accelerators is based on Siemens PLCs and on PVSS SCADA. After the transition from the LHC commissioning phase to its regular operation, there has been a number of additions and improvements to the vacuum control system. They were driven by new technical requirements and by feedback from the accelerator operators and vacuum specialists. New control functions have been implemented in the PLCs; new tools have been developed for the SCADA, while its ergonomic and navigation have been enhanced. (authors)

  17. Analysis of the steady-state operation of vacuum systems for fusion machines

    International Nuclear Information System (INIS)

    Roose, T.R.; Hoffman, M.A.; Carlson, G.A.

    1975-01-01

    A computer code named GASBAL was written to calculate the steady-state vacuum system performance of multi-chamber mirror machines as well as rather complex conventional multichamber vacuum systems. Application of the code, with some modifications, to the quasi-steady tokamak operating period should also be possible. Basically, GASBAL analyzes free molecular gas flow in a system consisting of a central chamber (the plasma chamber) connected by conductances to an arbitrary number of one- or two-chamber peripheral tanks. Each of the peripheral tanks may have vacuum pumping capability (pumping speed), sources of cold gas, and sources of energetic atoms. The central chamber may have actual vacuum pumping capability, as well as a plasma capable of ionizing injected atoms and impinging gas molecules and ''pumping'' them to a peripheral chamber. The GASBAL code was used in the preliminary design of a large mirror machine experiment--LLL's MX

  18. Ultra high vacuum system of the 3 MeV electron beam accelerator

    International Nuclear Information System (INIS)

    Puthran, G.P.; Jayaprakash, D.; Mishra, R.L.; Ghodke, S.R.; Majumder, R.; Mittal, K.C.; Sethi, R.C.

    2003-01-01

    Full text: A 3 MeV electron beam accelerator is coming up at the electron beam centre, Kharghar, Navi Mumbai. A vacuum of the order of 1x10 -7 mbar is desired in the beam line of the accelerator. The UHV system is spread over a height of 6 meters. The total surface area exposed to vacuum is 65,000 cm 2 and the volume is 200 litres. Distributed pumping is planned, to avoid undesirable vacuum gradient between any two sections of the beam-line. The electron beam is scanned in an area of 6 cms x 100 cms and it comes out of the scan-horn through a titanium foil of 50 micron thick. Hence the vacuum system is designed in such a way that, in the event of foil rupture during beam extraction, the electron gun, accelerating column and the pumps can be protected from sudden air rush. The vacuum in the beam-line can also be maintained in this condition. After changing the foil, scan-horn area can be separately pumped to bring the vacuum level as desired and can be opened to the beam-line. The design, vacuum pumping scheme and the safety aspects are discussed in this paper

  19. Engineering activities on the ITER representative diagnostic equatorial port plug

    International Nuclear Information System (INIS)

    Meunier, L.; Doceul, L.; Salasca, S.; Martins, J.-P.; Jullien, F.; Dechelle, Christian; Bidaud, Pierre; Pilard, Vincent; Terra, Alexis; Ogea, Mathieu; Ciattaglia, Emanuela; Walker, Christopher

    2009-01-01

    Most of ITER diagnostic systems are integrated in port plugs, which are water cooled stainless steel structures inserted into the vacuum vessel ports. The port plug must provide basic functions such as neutron and gamma shielding, supporting the first wall armour (BSM), closing the vacuum vessel ports, while supporting the diagnostic equipments. ITER diagnostic port plug must resist a severe environment like high temperature due to neutron interaction with the structures and high electromechanical loading during disruptions events. CEA has contributed to the design and integration tasks in the frame of the representative equatorial port plug EQ no. 01, in particular on the engineering, structural and thermal finite element analysis. These detailed analyses have highlighted some design issues which were worked out through different solutions. This paper contains a description of the engineering activities performed such as: -The static mechanical calculations of the top plate closure system under disruption load. -The static mechanical calculations of the BSM attachment to the port plug. These two first studies led to design changes proposals which significantly improved the behaviour of the structures but also showed that the safety margin with respect to design limits is quite low. -The design of a Diagnostic Shield Module (DSM) integrated inside the port plug and a proposition of attachment scheme, with respect to disruption loads. The manufacturing of the DSM has been taken into account, as well as diagnostic integration inside the structure and maintenance aspects. -The thermal assessment of the port plug under neutronic load during normal operation, with the optimization of the cooling system. The maximum temperature calculated in normal operation has been reduced from 900 deg. C to less than 400 deg. C in the front plate; and the cooling arrangement at the back of the port plug has been simplified without important temperature increase.

  20. Engineering activities on the ITER representative diagnostic equatorial port plug

    Energy Technology Data Exchange (ETDEWEB)

    Meunier, L. [Association Euratom CEA, CEA/DSM/IRFM (France)], E-mail: lmeunier@cea.fr; Doceul, L.; Salasca, S.; Martins, J.-P.; Jullien, F.; Dechelle, Christian; Bidaud, Pierre; Pilard, Vincent; Terra, Alexis; Ogea, Mathieu [Association Euratom CEA, CEA/DSM/IRFM (France); Ciattaglia, Emanuela [EFDA CSU, Garching (Germany); Walker, Christopher [ITER International Organisation (France)

    2009-06-15

    Most of ITER diagnostic systems are integrated in port plugs, which are water cooled stainless steel structures inserted into the vacuum vessel ports. The port plug must provide basic functions such as neutron and gamma shielding, supporting the first wall armour (BSM), closing the vacuum vessel ports, while supporting the diagnostic equipments. ITER diagnostic port plug must resist a severe environment like high temperature due to neutron interaction with the structures and high electromechanical loading during disruptions events. CEA has contributed to the design and integration tasks in the frame of the representative equatorial port plug EQ no. 01, in particular on the engineering, structural and thermal finite element analysis. These detailed analyses have highlighted some design issues which were worked out through different solutions. This paper contains a description of the engineering activities performed such as: -The static mechanical calculations of the top plate closure system under disruption load. -The static mechanical calculations of the BSM attachment to the port plug. These two first studies led to design changes proposals which significantly improved the behaviour of the structures but also showed that the safety margin with respect to design limits is quite low. -The design of a Diagnostic Shield Module (DSM) integrated inside the port plug and a proposition of attachment scheme, with respect to disruption loads. The manufacturing of the DSM has been taken into account, as well as diagnostic integration inside the structure and maintenance aspects. -The thermal assessment of the port plug under neutronic load during normal operation, with the optimization of the cooling system. The maximum temperature calculated in normal operation has been reduced from 900 deg. C to less than 400 deg. C in the front plate; and the cooling arrangement at the back of the port plug has been simplified without important temperature increase.