WorldWideScience

Sample records for main iter components

  1. Three-dimensional tolerance investigation on main ITER components

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  2. ITER plasma facing components

    International Nuclear Information System (INIS)

    Kuroda, T.; Vieider, G.; Akiba, M.

    1991-01-01

    This document summarizes results of the Conceptual Design Activities (1988-1990) for the International Thermonuclear Experimental Reactor (ITER) project, namely those that pertain to the plasma facing components of the reactor vessel, of which the main components are the first wall and the divertor plates. After an introduction and an executive summary, the principal functions of the plasma-facing components are delineated, i.e., (i) define the low-impurity region within which the plasma is produced, (ii) absorb the electromagnetic radiation and charged-particle flux from the plasma, and (iii) protect the blanket/shield components from the plasma. A list of critical design issues for the divertor plates and the first wall is given, followed by discussions of the divertor plate design (including the issues of material selection, erosion lifetime, design concepts, thermal and mechanical analysis, operating limits and overall lifetime, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, and advanced divertor concepts) and the first wall design (armor material and design, erosion lifetime, overall design concepts, thermal and mechanical analysis, lifetime and operating limits, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, an alternative first wall design, and the limiters used instead of the divertor plates during start-up). Refs, figs and tabs

  3. ITER plasma facing components, design and development

    International Nuclear Information System (INIS)

    Vieider, G.; Cardella, A.; Akiba, M.; Matera, R.; Watson, R.

    1991-01-01

    The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Plasma Facing Components (PFC) which focused on the following main tasks: (a) The definition of basic design concepts for the First Wall (FW) and Divertor Plates (DP), (b) the analysis of the performance and likely lifetime of these PFC designs including the identification of major critical issues, (c) the start of R and D work giving already first results, and the definition of the required further R and D program to support the contemplated ITER Engineering Design Activity (EDA). From the ITER CDA effort on PFC it is mainly concluded that: (a) The expected PFC operating conditions lead to design solutions at the limit of present technology in particular for the divertor, which may constrain the overall machine performance, (b) the development of convincing PFC designs requires an intensified R and D effort both on PFC technology and plasma physics. (orig.)

  4. Overview and status of ITER internal components

    International Nuclear Information System (INIS)

    Merola, Mario; Escourbiac, Frederic; Raffray, René; Chappuis, Philippe; Hirai, Takeshi; Martin, Alex

    2014-01-01

    Highlights: • Manufacturing technologies for the ITER internal components have been developed. • The Blanket System successfully went through its Final Design Review in April 2013. • The decision to start operation with a Divertor with a full-W armour has been taken. - Abstract: The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R and D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine

  5. Overview and status of ITER internal components

    Energy Technology Data Exchange (ETDEWEB)

    Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, René; Chappuis, Philippe; Hirai, Takeshi; Martin, Alex

    2014-10-15

    Highlights: • Manufacturing technologies for the ITER internal components have been developed. • The Blanket System successfully went through its Final Design Review in April 2013. • The decision to start operation with a Divertor with a full-W armour has been taken. - Abstract: The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R and D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine.

  6. Transport, logistics and packaging of ITER components

    International Nuclear Information System (INIS)

    Guerin, Olivier; Couturier, Bruno; Maas, Akko

    2005-01-01

    Cadarache, the European site for ITER, is located at around 50km as the crow flies from the sea. The feasibility of the transport of large and heavy ITER components has thus been thoroughly studied. These studies have covered the following items: - possible itineraries between the most convenient harbour (Fos) and Cadarache; - packaging (in particular for the largest and heaviest components); - means of transport (two types of trailers allowing to avoid lifting and load transfers); - logistics (analysis of transfer kinematics, including temporary storage); - administrative procedures and planning for the road adaptation, taking benefit of the recent successful implementation in the south-west of France of an itinerary for the Airbus A380 components. These studies, performed between 2001 and 2003, led to a viable solution, with a reasonable cost, fully supported by the French authorities. The planning necessary to implement the road modifications is also fully compatible with the expected dates of ITER components delivery

  7. Beryllium application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.; Barabash, V.; Cardella, A.; Jakeman, R.; Ioki, K.; Janeschitz, G.; Parker, R.; Tivey, R.; Pacher, H.D.; Wu, C.H.; Bartels, H.W.

    1997-01-01

    Beryllium is a candidate armour material for the in-vessel components of the International Thermonuclear Experimental Reactor (ITER), namely the primary first wall, the limiter, the baffle and the divertor. However, a number of issues arising from the performance requirements of the ITER plasma facing components (PFCs) must be addressed to better assess the attractiveness of Be as armour for these different components. These issues include heat loading limits arising from temperature and stress constraints under steady state conditions, armour lifetime including the effects of sputtering erosion as well as vaporisation and loss of melt during disruption events, tritium retention and permeation, and chemical hazards, in particular with respect to potential Be/steam reaction. Other issues such as fabrication and the possibility of in-situ repair are not performance-dependent but have an important impact on the overall assessment of Be as PFC armour. This paper describes the present view on Be application for ITER PFCs. The key issues are discussed including an assessment of the current level of understanding based on analysis and experimental data; and on-going activities as part of the ITER EDA R and D program are highlighted. (orig.)

  8. European technology activities to prepare for ITER component procurement

    International Nuclear Information System (INIS)

    Gasparotto, M.

    2006-01-01

    phase. In particular the manufacturing feasibility of the superconducting strand for the toroidal and poloidal field coils and of the toroidal field coils has been demonstrated, Studies of manufacturing techniques for the vacuum vessel, blanket modules and divertor are also in progress, and a number of EU industries have been prepared to successfully participate to the ITER construction. The paper will report on the main R(and)D activities performed in the EU and the major achievements in preparation for ITER component procurement. (author)

  9. Guidelines for Remote Handling Maintenance of ITER Neutral Beam Components

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  10. Carbon fiber composites application in ITER plasma facing components

    Science.gov (United States)

    Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

    1998-10-01

    Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

  11. Carbon fiber composites application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Matera, R.; Akiba, M.; Nakamura, K.; Bonal, J.P.; Pacher, H.D.; Roedig, M.; Vieider, G.; Wu, C.H.

    1998-01-01

    Carbon fiber composites (CFCs) are one of the candidate armour materials for the plasma facing components of the international thermonuclear experimental reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R and D needs are critically discussed. (orig.)

  12. Technologies for ITER divertor vertical target plasma facing components

    International Nuclear Information System (INIS)

    Schlosser, J.; Escourbiac, F.; Merola, M.; Fouquet, S.; Bayetti, P.; Cordier, J.J.; Grosman, A.; Missirlian, M.; Tivey, R.; Roedig, M.

    2005-01-01

    The ITER divertor vertical target has to sustain heat fluxes up to 20 MW m -2 . The concept developed for this plasma facing component working at steady state is based on carbon fibre composite armour for the lower straight part and tungsten for the curved upper part. The main challenges involved in the use of such components include the removal of the high heat fluxes deposited and mechanically and thermally joining the armour to the metallic heat sink, despite the mismatch in the thermal expansions. Two solutions based on the use of a CuCrZr hardened copper alloy and an active metal casting (AMC (registered) ) process were investigated during the ITER EDA phase: the first one called 'flat tile geometry' was mainly developed for the Tore Supra pumped limiter, the second one called 'monoblock geometry' was developed by the EU Participating Team for the ITER project. This paper presents a review of these two solutions and analyses their assets and drawbacks: pressure drop, critical heat flux, surface temperature and expected behaviour during operation, risks during the manufacture, control of the armour defects during the manufacture and at the reception, and the possibility of repairing defective tiles

  13. A new visible spectroscopy diagnostic for the JET ITER-like wall main chamber

    OpenAIRE

    Maggi, C. F.; Brezinsek, S.; Zastrow, K.-D.; JET-EFDA Contributors; Stamp, M. F.; Griph, S.; Heesterman, P.; Hogben, C.; Horton, A.; Meigs, A.; Morlock, C.; Studholme, W.

    2012-01-01

    In preparation for ITER, JET has been upgraded with a new ITER-like wall (ILW), whereby the main plasma facing components, previously of carbon, have been replaced by mainly Be in the main chamber and W in the divertor. As part of the many diagnostic enhancements, a new, survey, visible spectroscopy diagnostic has been installed for the characterization of the ILW. An array of eight lines-of-sight (LOS) view radially one of the two JET neutral beam shine through areas (W coated carbon fibre c...

  14. Iterative Mixture Component Pruning Algorithm for Gaussian Mixture PHD Filter

    Directory of Open Access Journals (Sweden)

    Xiaoxi Yan

    2014-01-01

    Full Text Available As far as the increasing number of mixture components in the Gaussian mixture PHD filter is concerned, an iterative mixture component pruning algorithm is proposed. The pruning algorithm is based on maximizing the posterior probability density of the mixture weights. The entropy distribution of the mixture weights is adopted as the prior distribution of mixture component parameters. The iterative update formulations of the mixture weights are derived by Lagrange multiplier and Lambert W function. Mixture components, whose weights become negative during iterative procedure, are pruned by setting corresponding mixture weights to zeros. In addition, multiple mixture components with similar parameters describing the same PHD peak can be merged into one mixture component in the algorithm. Simulation results show that the proposed iterative mixture component pruning algorithm is superior to the typical pruning algorithm based on thresholds.

  15. 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).

  16. Technical meeting on materials for in-vessel components of ITER

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.

    2000-01-01

    The Technical meeting on materials for in-vessel components of ITER was held at the ITER Joint Work Site in Garching from 31 January to 4 February. The main objectives of the meetings were: 1. to summarize the requirements, 2. to review new data, 3. to discuss in detail the R and D program and to discuss the material assessment report

  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. Design standard issues for ITER in-vessel components

    International Nuclear Information System (INIS)

    Majumdar, S.

    1994-01-01

    Unique requirements that must be addressed by a structural design code for the ITER have been summarized. Existing codes such as ASME Section III, or the French RCC-MR were developed primarily for fission reactor out-of-core components and are not directly applicable to the ITER. They may be used either as a guide for developing a design code for the ITER or as interim standards. However, new rules will be needed for handling the irradiation-induced embrittlement problems faced by the ITER blanket components. Design standards developed in the past for the design of fission reactor core components in the United States can be used as guides in this area

  19. Main maintenance operations for Test Blanket Systems in ITER TBM port cells

    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); Cortes, P.; Friconneau, J.-P.; Giancarli, L.M.; Gotewal, K.K.; Iseli, M.; Kim, B.Y.; Levesy, B.; Martins, J.-P.; Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Siarras, A. [Sogetti, Parc de la Duranne, 13857 Aix-en-Provence (France); Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues.

  20. Integration of remote refurbishment performed on ITER components

    Energy Technology Data Exchange (ETDEWEB)

    Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Antola, L. [AMEC, 31 Parc du Golf, CS 90519, 13596 Aix en Provence (France); Beaudoin, V. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Dremel, C. [Westinghouse, Electrique France/Astare, 122 Avenue de Hambourg, 13008 Marseille (France); Evrard, D. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Friconneau, J.P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Levesy, B.; Pitcher, C.S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2015-10-15

    Highlights: • System engineering approach to consolidate requirements to modify the layout of the Hot Cell. • Illustration of the loop between requirement and design. • Verification process. - Abstract: Internal components of the ITER Tokamak are replaced and transferred to the Hot Cell by remote handling equipment. These components include port plugs, cryopumps, divertor cassettes, blanket modules, etc. They are brought to the refurbishment area of the ITER Hot Cell Building for cleaning and maintenance, using remote handling techniques. The ITER refurbishment area will be unique in the world, when considering combination of size, quantity of complex component to refurbish in presence of radiation, activated dust and tritium. The refurbishment process to integrate covers a number of workstations to perform specific remote operations fully covered by a mast on crane system. This paper describes the integration of the Refurbishment Area, explaining the functions, the methodology followed, some illustrations of trade-off and safety improvements.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

    Design of the NB ports including duct liners under heat loads of the neutral beams has been developed. Design of the in-wall shielding has been developed in more details considering the supporting structure and the assembly method. The ferromagnetic inserts have previously not been installed in the outboard midplane region due to irregularity caused by the tangential ports for NB injection. Due to this configuration, the maximum ripple is relatively large (∝1 %) in a limited region of the plasma and the toroidal field flux lines fluctuate ∝10 mm in the FW region. To avoid these problems, additional ferromagnetic inserts are to be installed in the equatorial port region. Detailed studies were carried out on the ITER vacuum vessel to define appropriate codes and standards in the context of the ITER licensing in France. A set of draft documents regarding the ITER vacuum vessel structural code were prepared including an RCC-MR Addendum for the ITER VV with justified exceptions or modifications. The main deviation from the base Code is the extensive use of UT in lieu of radiography for the volumetric examination of all one-side access welds of the outer shell and field joint. The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. Development of new beryllium materials is progressing in China and Russia. The ITER limiters will be installed in equatorial ports at two toroidal locations. The limiter plasma-facing surface protrudes ∝8 cm from the FW during the start-up and shutdown phase. In the new limiter concept, the limiters are retracted by ∝8 cm during the plasma flat top phase. This concept gives important advantages; (i) mitigation of the particle and heat loads due to disruptions, ELMs and blobs, (ii) improvement of the power coupling with the ICRH antenna

  2. Development of radiation hardness components for ITER remote maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yagi, Toshiaki; Morita, Yousuke

    1998-04-01

    In the ITER, in-vessel remote handling is required to assemble and maintain in-vessel components in DT operations. Since in-vessel remote handling systems must operate under intense gamma ray radiation exceeding 30 kGy/h, their components must have sufficiently high radiation hardness to allow maintenance long enough in ITER in-vessel environments. Thus, extensive radiation tests and quality improvement, including optimization of material compositions, have been conducted through the ITER R and D program to develop radiation hardness components that meet radiation doses from 10 to 100 MGy at 10 kGy/h. This paper presents the latest on radiation hardness component development conducted by the Japan Home Team as a contribution to the ITER. The remote handling components tested are categorized for use in robotic or viewing systems, or as common components. Radiation tests have been conducted on commercially available products for screening, on modified products, and on new products to improve the radiation hardness. (author)

  3. Development of radiation hardness components for ITER remote maintenance

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; Ito, Akira; Yagi, Toshiaki; Morita, Yousuke

    1998-01-01

    In the ITER, in-vessel remote handling is required to assemble and maintain in-vessel components in DT operations. Since in-vessel remote handling systems must operate under intense gamma ray radiation exceeding 30 kGy/h, their components must have sufficiently high radiation hardness to allow maintenance long enough in ITER in-vessel environments. Thus, extensive radiation tests and quality improvement, including optimization of material compositions, have been conducted through the ITER R and D program to develop radiation hardness components that meet radiation doses from 10 to 100 MGy at 10 kGy/h. This paper presents the latest on radiation hardness component development conducted by the Japan Home Team as a contribution to the ITER. The remote handling components tested are categorized for use in robotic or viewing systems, or as common components. Radiation tests have been conducted on commercially available products for screening, on modified products, and on new products to improve the radiation hardness. (author)

  4. 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

  5. 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)

  6. F4E studies for the electromagnetic analysis of ITER components

    Energy Technology Data Exchange (ETDEWEB)

    Testoni, P., E-mail: pietro.testoni@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, c/ Josep Plá n.2, Barcelona (Spain); Cau, F.; Portone, A. [Fusion for Energy, Torres Diagonal Litoral B3, c/ Josep Plá n.2, Barcelona (Spain); Albanese, R. [Associazione EURATOM/ENEA/CREATE, DIETI, Università Federico II di Napoli, Napoli (Italy); Juirao, J. [Numerical Analysis TEChnologies S.L. (NATEC), c/ Marqués de San Esteban, 52 Entlo D Gijón (Spain)

    2014-10-15

    Highlights: • Several ITER components have been analyzed from the electromagnetic point of view. • Categorization of DINA load cases is described. • VDEs, MDs and MFD have been studied. • Integral values of forces and moments components versus time have been computed for all the ITER components under study. - Abstract: Fusion for Energy (F4E) is involved in a relevant number of activities in the area of electromagnetic analysis in support of ITER general design and EU in-kind procurement. In particular several ITER components (vacuum vessel, blanket shield modules and first wall panels, test blanket modules, ICRH antenna) are being analyzed from the electromagnetic point of view. In this paper we give an updated description of our main activities, highlighting the main assumptions, objectives, results and conclusions. The plasma instabilities we consider, typically disruptions and VDEs, can be both toroidally symmetric and asymmetric. This implies that, depending on the specific component and loading conditions, FE models we use span from a sector of 10 up to 360° of the ITER machine. The techniques for simulating the electromagnetic phenomena involved in a disruption and the postprocessing of the results to obtain the loads acting on the structures are described. Finally we summarize the typical loads applied to different components and give a critical view of the results.

  7. Status of R and D of the plasma facing components for the ITER divertor

    International Nuclear Information System (INIS)

    Mazul, I.V.; Akiba, M.; Arkhipov, I.

    2001-01-01

    The paper reports the progress made by the ITER Home Teams in the development of robust carbon and tungsten armoured plasma facing components for the ITER divertor. The activities on the development and study of armour materials, joining technologies, non-destructive evaluation techniques, high heat flux testing of manufactured components and neutron irradiation resistance studies are presented. The results of these activities confirm the feasibility of the main divertor components. Examples of the fruitful collaboration between Parties and future R and D needs are also described. (author)

  8. Main heat transfer components for SNR-300

    International Nuclear Information System (INIS)

    De Haas Van Dorsser, A.H.

    1976-01-01

    Early in the joint German-Belgium-Dutch fast breeder programme it was decided that all main components should be tested, if possible at full scale, before fabrication of the actual SNR-300 components. Descriptions are given of the results of testing, and subsequent modifications, of the pumps, intermediate heat exchangers, and steam generators. A full scale model of the primary pump, free surface vertical shaft centrifugal type, was constructed and tested in the 5000 cubic metres per hour pump test facility erected at Bensberg. A 70 MW model of an intermediate heat exchanger, straight tube type with floating head, was tested in the 50 MW steam generator test station at Hengelo. Also tested in the Hengelo facility was an almost full scale straight tube 50 MW steam generator and subsequently a 50 MW helical tube evaporator. The latter tests were of more than 3000 h operation and resulted in minor changes in design and manufacturing operation. (U.K.)

  9. Supply of a prototype component for the ITER divertor baffle

    International Nuclear Information System (INIS)

    Bobin-Vastra, I.; Febvre, M.; Schedler, B.; Ploechl, L.; Bouveret, Y.; Cauvin, D.; Raisson, G.; Merola, M.

    2001-01-01

    The ITER divertor baffle is one of the Plasma facing components which are developed in the frame of the ITER concept. The supply consisted in the manufacturing of four panels with four First Wall geometries using macroblock or heat sink+armour concepts. DS-Copper, and CuCrZr were the materials for the heat sink, and CFC or Tungsten Plasma spray were the armour. The panels included two Copper-based tubes each. The final purpose is the comparison of the fabricability of each type and the performances of each panel under heat fluxes

  10. Design, Manufacture, and Experimental Serviceability Validation of ITER Blanket Components

    Science.gov (United States)

    Leshukov, A. Yu.; Strebkov, Yu. S.; Sviridenko, M. N.; Safronov, V. M.; Putrik, A. B.

    2017-12-01

    In 2014, the Russian Federation and the ITER International Organization signed two Procurement Arrangements (PAs) for ITER blanket components: 1.6.P1ARF.01 "Blanket First Wall" of February 14, 2014, and 1.6.P3.RF.01 "Blanket Module Connections" of December 19, 2014. The first PA stipulates development, manufacture, testing, and delivery to the ITER site of 179 Enhanced Heat Flux (EHF) First Wall (FW) Panels intended for withstanding the heat flux from the plasma up to 4.7MW/m2. Two Russian institutions, NIIEFA (Efremov Institute) and NIKIET, are responsible for the implementation of this PA. NIIEFA manufactures plasma-facing components (PFCs) of the EHF FW panels and performs the final assembly and testing of the panels, and NIKIET manufactures FW beam structures, load-bearing structures of PFCs, and all elements of the panel attachment system. As for the second PA, NIKIET is the sole official supplier of flexible blanket supports, electrical insulation key pads (EIKPs), and blanket module/vacuum vessel electrical connectors. Joint activities of NIKIET and NIIEFA for implementing PA 1.6.P1ARF.01 are briefly described, and information on implementation of PA 1.6.P3.RF.01 is given. Results of the engineering design and research efforts in the scope of the above PAs in 2015-2016 are reported, and results of developing the technology for manufacturing ITER blanket components are presented.

  11. 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.

  12. 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)

  13. Damage evaluation under thermal fatigue of a vertical target full scale component for the ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Merola, M.; Durocher, A.; Bobin-Vastra, I.; Schedler, B.

    2007-01-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a Full Scale Vertical Target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses thermographic examination and thermal fatigue testing results obtained on this component. The study includes thermal analysis, with a tentative proposal to evaluate with finite element approach the location/size of defects and the possible propagation during fatigue cycling

  14. Examination of high heat flux components for the ITER divertor after thermal fatigue testing

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Schmidt, A.; Riccardi, B.; Bobin-Vastra, I.

    2011-01-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour-heat sink interface.

  15. Examination of high heat flux components for the ITER divertor after thermal fatigue testing

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M., E-mail: marc.missirlian@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Escourbiac, F., E-mail: frederic.escourbiac@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Schmidt, A., E-mail: a.schmidt@fz-juelich.de [Forschungszentrum Juelich, IFE-2 (Germany); Riccardi, B., E-mail: Bruno.Riccardi@f4e.europa.eu [Fusion For Energy, E-08019 Barcelona (Spain); Bobin-Vastra, I., E-mail: isabelle.bobinvastra@areva.com [AREVA-NP, 71200 Le Creusot (France)

    2011-10-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour-heat sink interface.

  16. Technological challenges at ITER plasma facing components production in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Mazul, I.V., E-mail: mazuliv@niiefa.spb.su [Efremov Institute, 196641 St. Petersburg (Russian Federation); Belyakov, V.A.; Gervash, A.A.; Giniyatulin, R.N.; Guryeva, T.M.; Kuznetsov, V.E.; Makhankov, A.N.; Okunev, A.A. [Efremov Institute, 196641 St. Petersburg (Russian Federation); Sevryukov, O.N. [MEPhI, 115409 Moscow (Russian Federation)

    2016-11-01

    Highlights: • Technological aspects of ITER PFC manufacturing in Russia are presented. • Range of technologies to be used during manufacturing of ITER PFC at Efremov Institute has been, in general, defined and their complexity, originality and difficulty are described. • Some features and challenges of welding, brazing and various tests are discussed. - Abstract: Major part of ITER plasma facing components will be manufactured in the Russian Federation (RF). Operational conditions and other requirements to these components, as well as the scale of production, are quite unique. These unique features and related technological solutions found in the frame of the project are discussed. Procedure breakdown and results of qualification for the proposed technologies and potential producers are presented, based on mockups production and testing. Design of qualification mockups and prototypes, testing programs and results are described. Basic quantitative and qualitative parameters of manufactured components and methods of quality control are presented. Critical manufacturing issues and prospects for unique production for future fusion needs are discussed.

  17. A new visible spectroscopy diagnostic for the JET ITER-like wall main chamber

    International Nuclear Information System (INIS)

    Maggi, C. F.; Brezinsek, S.; Stamp, M. F.; Griph, S.; Heesterman, P.; Hogben, C.; Horton, A.; Meigs, A.; Studholme, W.; Zastrow, K.-D.; Morlock, C.

    2012-01-01

    In preparation for ITER, JET has been upgraded with a new ITER-like wall (ILW), whereby the main plasma facing components, previously of carbon, have been replaced by mainly Be in the main chamber and W in the divertor. As part of the many diagnostic enhancements, a new, survey, visible spectroscopy diagnostic has been installed for the characterization of the ILW. An array of eight lines-of-sight (LOS) view radially one of the two JET neutral beam shine through areas (W coated carbon fibre composite tiles) at the inner wall. In addition, one vertical LOS views the solid W tile at the outer divertor. The light emitted from the plasma is coupled to a series of compact overview spectrometers, with overall wavelength range of 380–960 nm and to one high resolution Echelle overview spectrometer covering the wavelength range 365–720 nm. The new survey diagnostic has been absolutely calibrated in situ by means of a radiometric light source placed inside the JET vessel in front of the whole optical path and operated by remote handling. The diagnostic is operated in every JET discharge, routinely monitoring photon fluxes from intrinsic and extrinsic impurities (e.g., Be, C, W, N, and Ne), molecules (e.g., BeD, D 2 , ND) and main chamber and divertor recycling (typically Dα, Dβ, and Dγ). The paper presents a technical description of the diagnostic and first measurements during JET discharges.

  18. Armour Materials for the ITER Plasma Facing Components

    Science.gov (United States)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

    The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

  19. Armour materials for the ITER plasma facing components

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A.R.

    1999-01-01

    The selection of the armour materials for the plasma facing components (PFCs) of the international thermonuclear experimental reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-a-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R and D. (orig.)

  20. Design of the ITER Plasma-Facing Components

    Energy Technology Data Exchange (ETDEWEB)

    Merola, M.

    2009-07-01

    The ITER plasma-facing components cover an area of about 850 m{sup 2} and consist of the Divertor, the Blanket and the Test Blanket Modules (TBMs) with their corresponding frames. The Divertor is located at the bottom of the plasma chamber and is aimed at exhausting the major part of the plasma thermal power (including alpha power) and at minimizing the helium and impurity content in the plasma. It consists of 54 cassette assemblies. Each assembly has 3 plasma-facing components (PFCs), namely the inner and outer target and the dome, which are mounted onto a steel support structure, the cassette body. The targets directly intercept the magnetic field lines and are designed to withstand heat fluxes as high as 20 MW/m{sup 2}. CFC is the reference design solution for the armour of the lower part of the targets. However, the resultant high erosion rate could potentially limit machine operation in the DT phase (due to co-deposition with T). Therefore, prior to the DT phase, the divertor PFCs will be replaced with a new set entirely covered with W armour. The Divertor is a RH Class 1 component, which is planned to be replaced 3 times during the 20 years of the ITER operation. The construction phase of the ITER Divertor is being launched. The Blanket covers the largest fraction of the plasma-facing surface. Each of the 440 Blanket modules consists of a first wall (FW) panel, which is mechanically attached onto a Shield Module (SM). The design heat flux is set up to 1 or 5 MW/m{sup 2}. The FW panels are covered by Be tiles, which are joined onto a copper alloy (CuCrZr) heat sink, which is in turn intimately joined onto a 316L(N) stainless steel part. The SM is a block of 316L(N)-IG steel, where an array of cooling channels are obtained by machining and welding. The TBMs are mock-ups of DEMO breeding blankets. There are three ITER equatorial ports devoted to TBM testing, each of them allocating two TBMs, inserted in a thick steel frame. The frame is a water-cooled 316L

  1. Structural materials for ITER in-vessel component design

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

  2. Interaction between main components in wind farms

    DEFF Research Database (Denmark)

    Holdyk, Andrzej; Koldby, Erik

    and the simplicity of the measurement methods using the device makes it a good candidate for performing black-box modelling of multiports whenever such models are not available from the manufacturers. Parametric variation method developed for EMT simulations in ATP-EMTP is a good tool for performing large...... with Frequency Domain Severity Factor proved to be a robust tool in assessing stresses on electric components arising from transient phenomena in offshore wind farms, including the voltage magnitude and frequency of oscillations. Quarter-wave resonance frequency is a good approximation of resonance frequency...... as well as performing parametric variation studies. Methods and tools were developed and shown to perform and estimate the severity of a potential mid- and high- frequency interaction between electric components in OWFs by robust sensitivity analysis in commercial EMT simulation tool. Performing...

  3. Main components of business cards design

    Directory of Open Access Journals (Sweden)

    Ю. В. Романенкова

    2003-03-01

    Full Text Available The essay is dedicated to the urgent problem of necessity of creation of professional design of business cards, that are important part of the image of modem businessman. There are classification of cards by functional principle, the functions of cards of each type were analyzed. All components of business card, variants of its composition schemes, color characteristics, principles of use of trade marks and other design elements have been allocated

  4. Magnetic field compatibility of components for ITER electrical distribution systems

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  5. 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)

  6. 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)

  7. 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

  8. Design, manufacture and initial operation of the beryllium components of the JET ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Riccardo, V., E-mail: valeria.riccardo@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Lomas, P.; Matthews, G.F. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Nunes, I. [Associação EURATOM-IST, IPFN – Laboratório Associado, IST, Lisbon (Portugal); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Thompson, V. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Villedieu, E. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

    2013-10-15

    Highlights: ► 40 m{sup 2} of plasma facing surface covered with bulk Be re-using existing supports, designed for C-based tiles (hence for much lower disruption loads). ► Optimization of power handling to allow compatibility with higher (×1.5) and longer (×2) neutral beam power. ► Beryllium re-cycling. ► Machining and cleaning to ultra high vacuum standards of <350 μm thin castellations in Be. ► Quality control to minimize installation problems (proto-types, full scale jigs, inspections). -- Abstract: The aim of the JET ITER-like wall project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc.) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets.

  9. Design, manufacture and initial operation of the beryllium components of the JET ITER-like wall

    International Nuclear Information System (INIS)

    Riccardo, V.; Lomas, P.; Matthews, G.F.; Nunes, I.; Thompson, V.; Villedieu, E.

    2013-01-01

    Highlights: ► 40 m 2 of plasma facing surface covered with bulk Be re-using existing supports, designed for C-based tiles (hence for much lower disruption loads). ► Optimization of power handling to allow compatibility with higher (×1.5) and longer (×2) neutral beam power. ► Beryllium re-cycling. ► Machining and cleaning to ultra high vacuum standards of <350 μm thin castellations in Be. ► Quality control to minimize installation problems (proto-types, full scale jigs, inspections). -- Abstract: The aim of the JET ITER-like wall project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc.) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets

  10. Joining technologies for the plasma facing components of ITER

    International Nuclear Information System (INIS)

    Barabash, V.; Kalinin, G.; Matera, R.

    1998-01-01

    An extensive R and D program on the development of the joining technologies between armour (beryllium, tungsten and carbon fibre composites)/copper alloys heat sink and copper alloys/ stainless steel has been carried out by ITER Home Teams. A brief review of this R and D program is presented in this paper. Based on the results, reference technologies for use in ITER have been selected and recommended for further development. (author)

  11. Main components and content of sports volunteer activities

    OpenAIRE

    Петренко, Ірина

    2017-01-01

    Iryna PetrenkоPurpose: identification of the main structural components and content of sports volunteer activities. Material & Methods: used analysis of literature and documents, organizational analysis. Result: basic structural components of sports volunteer activity are defined. The content of sports volunteer activity is disclosed. Conclusion: sports volunteer activity includes the following structural components: subject, object, purpose, motivation, means, actions; subject is a sport...

  12. Manufacturing studies of double wall components for the ITER EC H and CD upper launcher

    International Nuclear Information System (INIS)

    Spaeh, P.; Aiello, G.; Goldmann, A.; Kleefeldt, K.; Kroiss, A.; Meier, A.; Obermeier, C.; Scherer, T.; Schreck, S.; Serikov, A.; Strauss, D.; Vaccaro, A.

    2012-01-01

    Highlights: ► Double wall manufacturing technologies for ITER In-vessel components. ► Rigid and safe accommodation of ECRH heating and current drive systems. ► Thermo hydraulic analysis of coolant flow in double-wall structures. - Abstract: To counteract plasma instabilities, Electron Cyclotron Launchers will be installed in four of the ITER Upper Ports. The structural system of an EC Upper Launcher accommodates the MM-wave-components and has to meet strong demands on alignment, removal of nuclear heat loads, mechanical strength and nuclear shielding. The EC Upper Launcher has successfully undergone the Preliminary Design Review in 2009 and is now in the final design phase. Nuclear heat loads from 0.1 W/cm 3 up to 0.8 W/cm 3 will affect the front area of the launcher main frame. To guarantee save and homogenous removal of those heat loads, the front part of the launcher main frame is designed as a double wall steel-casing with cooling channels inside the shell structure. To finalize the design of this double wall component, the main emphasis is now to define the cooling channels geometry and to identify the optimum manufacturing route to assure adequate flow of coolant and sufficient mechanical strength in compliance with required dimension tolerances and quality of the welds. Several manufacturing options have been investigated and were evaluated by computational analysis and fabrication of pre-prototypes. To come to a final design, the most promising route will be chosen to manufacture a full-size mock-up of the double wall main frame. It will be tested at the KIT Launcher Handling Test facility to check the compliance with the design goals related to geometrical accuracy and thermo-hydraulic characteristics. This paper describes the design and the manufacturing routes of the prototypic double wall main frame.

  13. Manufacturing studies of double wall components for the ITER EC H and CD upper launcher

    Energy Technology Data Exchange (ETDEWEB)

    Spaeh, P., E-mail: peter.spaeh@kit.edu [Institute for Applied Materials, Karlsruhe Institute of Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, G. [Institute for Applied Materials, Karlsruhe Institute of Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany); Goldmann, A. [MAN Diesel and Turbo, D-94452 Deggendorf, P.O. Box 3640, D-76021 Karlsruhe (Germany); Kleefeldt, K. [Institute for Applied Materials, Karlsruhe Institute of Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany); Kroiss, A. [MAN Diesel and Turbo, D-94452 Deggendorf, P.O. Box 3640, D-76021 Karlsruhe (Germany); Meier, A. [Institute for Applied Materials, Karlsruhe Institute of Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany); Obermeier, C. [MAN Diesel and Turbo, D-94452 Deggendorf, P.O. Box 3640, D-76021 Karlsruhe (Germany); Scherer, T.; Schreck, S. [Institute for Applied Materials, Karlsruhe Institute of Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany); Serikov, A. [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany); Strauss, D.; Vaccaro, A. [Institute for Applied Materials, Karlsruhe Institute of Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Double wall manufacturing technologies for ITER In-vessel components. Black-Right-Pointing-Pointer Rigid and safe accommodation of ECRH heating and current drive systems. Black-Right-Pointing-Pointer Thermo hydraulic analysis of coolant flow in double-wall structures. - Abstract: To counteract plasma instabilities, Electron Cyclotron Launchers will be installed in four of the ITER Upper Ports. The structural system of an EC Upper Launcher accommodates the MM-wave-components and has to meet strong demands on alignment, removal of nuclear heat loads, mechanical strength and nuclear shielding. The EC Upper Launcher has successfully undergone the Preliminary Design Review in 2009 and is now in the final design phase. Nuclear heat loads from 0.1 W/cm{sup 3} up to 0.8 W/cm{sup 3} will affect the front area of the launcher main frame. To guarantee save and homogenous removal of those heat loads, the front part of the launcher main frame is designed as a double wall steel-casing with cooling channels inside the shell structure. To finalize the design of this double wall component, the main emphasis is now to define the cooling channels geometry and to identify the optimum manufacturing route to assure adequate flow of coolant and sufficient mechanical strength in compliance with required dimension tolerances and quality of the welds. Several manufacturing options have been investigated and were evaluated by computational analysis and fabrication of pre-prototypes. To come to a final design, the most promising route will be chosen to manufacture a full-size mock-up of the double wall main frame. It will be tested at the KIT Launcher Handling Test facility to check the compliance with the design goals related to geometrical accuracy and thermo-hydraulic characteristics. This paper describes the design and the manufacturing routes of the prototypic double wall main frame.

  14. Testing of high heat flux components manufactured by ENEA for ITER divertor

    International Nuclear Information System (INIS)

    Visca, Eliseo; Escourbiac, F.; Libera, S.; Mancini, A.; Mazzone, G.; Merola, M.; Pizzuto, A.

    2009-01-01

    ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R and D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested. A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process. The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France). The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m 2 , 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m 2 , 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m 2 Critical Heat Flux was measured at relevant thermal-hydraulics conditions at the end of the testing campaign. This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up. These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).

  15. Principal components and iterative regression analysis of geophysical series: Application to Sunspot number (1750 2004)

    Science.gov (United States)

    Nordemann, D. J. R.; Rigozo, N. R.; de Souza Echer, M. P.; Echer, E.

    2008-11-01

    We present here an implementation of a least squares iterative regression method applied to the sine functions embedded in the principal components extracted from geophysical time series. This method seems to represent a useful improvement for the non-stationary time series periodicity quantitative analysis. The principal components determination followed by the least squares iterative regression method was implemented in an algorithm written in the Scilab (2006) language. The main result of the method is to obtain the set of sine functions embedded in the series analyzed in decreasing order of significance, from the most important ones, likely to represent the physical processes involved in the generation of the series, to the less important ones that represent noise components. Taking into account the need of a deeper knowledge of the Sun's past history and its implication to global climate change, the method was applied to the Sunspot Number series (1750-2004). With the threshold and parameter values used here, the application of the method leads to a total of 441 explicit sine functions, among which 65 were considered as being significant and were used for a reconstruction that gave a normalized mean squared error of 0.146.

  16. The heat removal capability of actively cooled plasma-facing components for the ITER divertor

    Science.gov (United States)

    Missirlian, M.; Richou, M.; Riccardi, B.; Gavila, P.; Loarer, T.; Constans, S.

    2011-12-01

    Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m-2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m-2 for the CFC-armoured tiles and 15 MW m-2 for the W-armoured tiles, respectively.

  17. The heat removal capability of actively cooled plasma-facing components for the ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M; Richou, M; Loarer, T; Riccardi, B; Gavila, P; Constans, S

    2011-01-01

    Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m - 2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m - 2 for the CFC-armoured tiles and 15 MW m - 2 for the W-armoured tiles, respectively.

  18. The main objectives of lifetime management of reactor unit components

    International Nuclear Information System (INIS)

    Dragunov, Y.; Kurakov, Y.

    1998-01-01

    The main objectives of the work concerned with life management of reactor components in Russian Federation are as follows: development of regulations in the field of NPP components ageing and lifetime management; investigations of ageing processes; residual life evaluation taking into account the actual state of NPP systems, real loading conditions and number of load cycles, results of in-service inspections; development and implementation of measures for maintaining/enhancing the NPP safety

  19. European development of carbon armoured plasma facing components for ITER

    International Nuclear Information System (INIS)

    Merola, M.; Vieider, G.; Wu, C.; Schedler, B.; Chappuis, P.; Escourbiac, F.; Schlosser, J.; Duwe, R.; Roedig, M.; Febvre, M.; Grattarola, M.; Tahtinen, S.; Vesprini, R.

    2001-01-01

    After a brief description of the rationale of the material and geometry selection for each carbon armoured plasma facing components, this paper describes the European development of the two basic geometries, namely the monoblock and the flat tile. An overview of the non-destructive inspection techniques specifically developed for these components is also presented. (orig.)

  20. 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.

  1. Effects of asymmetric vertical disruptions on ITER components

    International Nuclear Information System (INIS)

    Albanese, R.; Carpentieri, B.; Cavinato, M.; Minucci, S.; Palmaccio, R.; Portone, A.; Rubinacci, G.; Testoni, P.; Ventre, S.; Villone, F.

    2015-01-01

    Highlights: • Halo current analysis of AVDEs (asymmetric VDEs) is performed. • Both resistive and inductive effects are considered. • Suitable compression techniques and supercomputing resources are used. • The vertical force on the sectors is nearly uniform. • The radial loads on the various sectors are very different. - Abstract: This paper deals with the halo current distribution due to asymmetric vertical displacement events (VDEs) and the subsequent force distributions on the conducting structures in the ITER tokamak. Both the eddy and halo current analyses have been carried out using the 3D code CARIDDI, based on an integral formulation in the conducting region. The plasma plays the role of a source term. The axisymmetric time evolution of the plasma is taken by 2D axisymmetric simulations. The most critical case is a slow VDE downward combined with an n = 1 kink, which may yield large horizontal forces and peaking factors. A simplified n = 1, m = 1 kink model is taken, given by a rigid horizontal displacement accompanied by a tilt. The halo currents are treated as injected currents on the faces of the first wall hit by the plasma. To take into account the inductive effects, which are important especially in the transient phases, suitable compression techniques and supercomputing resources have been utilized. In the worst case the total vertical force on the structure due to the halo currents is about 90 MN downwards (about 30 of which on the divertor); the horizontal force is about 4 MN (about half of which on the divertor); the distribution of the vertical force on the sectors is nearly uniform, whereas the radial loads on the various sectors are very different from each other

  2. Effects of asymmetric vertical disruptions on ITER components

    Energy Technology Data Exchange (ETDEWEB)

    Albanese, R. [Associazione EURATOM/ENEA/CREATE, DIETI, Università di Napoli Federico II, Napoli (Italy); Carpentieri, B. [Johann Bernoulli Institute for Mathematics and Computer Science, University of Groningen, Groningen (Netherlands); Cavinato, M. [Fusion for Energy, Torres Diagonal Litoral B3, c/ Josep Plá n.2, Barcelona (Spain); Minucci, S. [Associazione EURATOM/ENEA/CREATE, DIETI, Università di Napoli Federico II, Napoli (Italy); Palmaccio, R. [Associazione EURATOM/ENEA/CREATE, DIEI, Università di Cassino e del Lazio Meridionale, Cassino, FR (Italy); Portone, A. [Fusion for Energy, Torres Diagonal Litoral B3, c/ Josep Plá n.2, Barcelona (Spain); Rubinacci, G. [Associazione EURATOM/ENEA/CREATE, DIETI, Università di Napoli Federico II, Napoli (Italy); Testoni, P., E-mail: pietro.testoni@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, c/ Josep Plá n.2, Barcelona (Spain); Ventre, S.; Villone, F. [Associazione EURATOM/ENEA/CREATE, DIEI, Università di Cassino e del Lazio Meridionale, Cassino, FR (Italy)

    2015-05-15

    Highlights: • Halo current analysis of AVDEs (asymmetric VDEs) is performed. • Both resistive and inductive effects are considered. • Suitable compression techniques and supercomputing resources are used. • The vertical force on the sectors is nearly uniform. • The radial loads on the various sectors are very different. - Abstract: This paper deals with the halo current distribution due to asymmetric vertical displacement events (VDEs) and the subsequent force distributions on the conducting structures in the ITER tokamak. Both the eddy and halo current analyses have been carried out using the 3D code CARIDDI, based on an integral formulation in the conducting region. The plasma plays the role of a source term. The axisymmetric time evolution of the plasma is taken by 2D axisymmetric simulations. The most critical case is a slow VDE downward combined with an n = 1 kink, which may yield large horizontal forces and peaking factors. A simplified n = 1, m = 1 kink model is taken, given by a rigid horizontal displacement accompanied by a tilt. The halo currents are treated as injected currents on the faces of the first wall hit by the plasma. To take into account the inductive effects, which are important especially in the transient phases, suitable compression techniques and supercomputing resources have been utilized. In the worst case the total vertical force on the structure due to the halo currents is about 90 MN downwards (about 30 of which on the divertor); the horizontal force is about 4 MN (about half of which on the divertor); the distribution of the vertical force on the sectors is nearly uniform, whereas the radial loads on the various sectors are very different from each other.

  3. Iter

    Science.gov (United States)

    Iotti, Robert

    2015-04-01

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

  4. Beryllium assessment and recommendation for application in ITER plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Barabash, V.; Tanaka, S.; Matera, R. [ITER Joint Central Team, Muenchen (Germany)

    1998-01-01

    The design status of the ITER Plasma Facing Components (PFC) is presented. The operational conditions of the armour material for the different components are summarized. Beryllium is the reference armour material for the Primary Wall, Baffle and Limiter and the back-up material for the Divertor Dome. The activities on the selection of the Be grades and the joining technologies are reviewed. (author)

  5. Component biomass equations for black spruce in Maine

    Science.gov (United States)

    M. M. Czapowskyj; D. J. Robison; R. D. Briggs; E. H. White; E. H. White

    1985-01-01

    Component biomass prediction equations are presented for young black spruce (Picea mariana B.S.P. (Mill,:)) in northern Maine. A weighted least squares model was used to construct the eq~iationsfo r small trees from 1 to 15 cm d.b.h., and an ordinary least squares model for trees less than 2 m in height. A linearized allometric model was also tested but was not used....

  6. 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.

  7. 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.)

  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. Effect of ITER components manufacturing cycle on the irradiation behaviour of 316L(N)-IG steel

    International Nuclear Information System (INIS)

    Rodchenkov, B.S.; Prokhorov, V.I.; Makarov, O.Yu.; Shamardin, V.K.; Kalinin, G.M.; Strebkov, Yu.S.; Golosov, O.A.

    2000-01-01

    The main options for the manufacturing of high heat flux (HHF) components is hot isostatic pressing (HIP) using either solid pieces or powder. There was no database on the radiation behaviour of these materials, and in particular stainless steel (SS) 316L(N)-IG with ITER components manufacturing thermal cycle. Irradiation of wrought steel, powder-HIP, solid-HIP and HIPed joints has been performed within the framework of an ITER task. Specimens cut from 316L(N)-IG plate, HIP products, and solid-HIP joints were irradiated in the SM-3 reactor in Dimitrovgrad up to 4 and 10 dpa at 175 deg. C and 265 deg. C. The paper describes the results of post-irradiation tensile and fracture toughness tests

  10. Main components and content of sports volunteer activities

    Directory of Open Access Journals (Sweden)

    Iryna Petrenkо

    2017-12-01

    Full Text Available Purpose: identification of the main structural components and content of sports volunteer activities. Material & Methods: used analysis of literature and documents, organizational analysis. Result: basic structural components of sports volunteer activity are defined. The content of sports volunteer activity is disclosed. Conclusion: sports volunteer activity includes the following structural components: subject, object, purpose, motivation, means, actions; subject is a sports volunteer, the object is a sports competition, the goal is to provide gratuitous assistance for a quality competition, the means are the special knowledge, skills, communication abilities of sports volunteers, actions should be understood as types of volunteer activities and functions that volunteers perform during the preparation and conduct of competitions. Main types of sports volunteer activity are: 1 organizational; 2 judiciary; 3 coaching; 4 legal; 5 medical. Functions that volunteers perform in the competition system are general and special. Content of the functions of sports volunteering depends on the specifics of the sports, the rank of the competition, the specifics of the competition for people with special needs.

  11. Calorimetry of the JET ITER-Like Wall components

    Energy Technology Data Exchange (ETDEWEB)

    Devaux, S., E-mail: stephane.devaux@ccfe.ac.uk [Max-Planck-Institut für Plasmaphysik, EURATOM-Assoziation, D-85748 Garching (Germany); Arnoux, G. [EURATOM/CCFE Association, Culham Science Center, Abingdon, Oxon OX14 3DB (United Kingdom); Corre, Y. [Association EURTATOM-CEA, CEA/DSM/IRFM, Cadarache, 13108 Saint Paul lez Durance (France); Gardarein, J-L.; Gaspar, J. [IUSTI UMR CNRS 7343, Aix-Marseille University Marseille (France); Jacquet, P. [Max-Planck-Institut für Plasmaphysik, EURATOM-Assoziation, D-85748 Garching (Germany); Marcotte, F. [EURATOM/CCFE Association, Culham Science Center, Abingdon, Oxon OX14 3DB (United Kingdom); Matthews, G.; Beaumont, P.; Cramp, S.; Dalley, S.; Kinna, D.; Horton, A.; Lomas, P. [Max-Planck-Institut für Plasmaphysik, EURATOM-Assoziation, D-85748 Garching (Germany); Mertens, Ph. [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich GmbH, Association EURATOM-FZJ, D-52425 Jülich (Germany); Riccardo, V. [Max-Planck-Institut für Plasmaphysik, EURATOM-Assoziation, D-85748 Garching (Germany); Valcàrcel, D. [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, P-1049-001 Lisboa (Portugal)

    2013-07-15

    As part of the ILW project, new diagnostics have been installed in order to protect the plasma-facing components (PFCs). Here we present the diagnostics used to monitor the PFC temperature, thermocouples and cameras, and assess the consistency of their measurements. In dedicated limited L-mode plasmas, the surface of the limiter tiles are heated up to 900 °C. The comparison of surface temperature measurements from IR and near IR cameras, which have been calibrated against a black-body source, leads to a Be emissivity of 0.18, comparable with the theoretical one. Energy calculation derived from thermocouples, which are embedded in both limiters and divertor target plates (W-coated CFC), is compared to a 1D model based on thermal quadrupole approach (benchmarked with an ANSYS model) associated to an inversion computation. The analysis of 20 pulses shows that a good energy balance is achieved within the error bar of the model, assessed to be of 30%.

  12. Calorimetry of the JET ITER-Like Wall components

    International Nuclear Information System (INIS)

    Devaux, S.; Arnoux, G.; Corre, Y.; Gardarein, J-L.; Gaspar, J.; Jacquet, P.; Marcotte, F.; Matthews, G.; Beaumont, P.; Cramp, S.; Dalley, S.; Kinna, D.; Horton, A.; Lomas, P.; Mertens, Ph.; Riccardo, V.; Valcàrcel, D.

    2013-01-01

    As part of the ILW project, new diagnostics have been installed in order to protect the plasma-facing components (PFCs). Here we present the diagnostics used to monitor the PFC temperature, thermocouples and cameras, and assess the consistency of their measurements. In dedicated limited L-mode plasmas, the surface of the limiter tiles are heated up to 900 °C. The comparison of surface temperature measurements from IR and near IR cameras, which have been calibrated against a black-body source, leads to a Be emissivity of 0.18, comparable with the theoretical one. Energy calculation derived from thermocouples, which are embedded in both limiters and divertor target plates (W-coated CFC), is compared to a 1D model based on thermal quadrupole approach (benchmarked with an ANSYS model) associated to an inversion computation. The analysis of 20 pulses shows that a good energy balance is achieved within the error bar of the model, assessed to be of 30%

  13. Smart Grids: short history, main components and perspectives

    International Nuclear Information System (INIS)

    Belhomme, R.; Maire, J.

    2010-01-01

    This paper tries to describe the smart grids in their entirety. In a first part, two definitions are given for the smart grids, the main drivers for their development are briefly described, along with an overview of the initiatives and projects in different parts of the world: Europe, USA, Asia and Pacific area. The second part is devoted to the main components of the smart grids. The following are considered: distributed generation and renewables, energy storage, demand side integration, intelligent buildings, smart meters, communication infrastructures, distribution and transmission networks, micro-grids, flexibility of the generation park, electric vehicles and regulatory issues. Finally, the conclusion gives a short discussion of some important issues, as well as of the benefits of demonstration projects. (authors)

  14. Myoepithelial cells are the main component in pleomorphic adenomas?

    Science.gov (United States)

    Ponce Bravo, Santa; Ledesma Montes, Constantino; López Becerril, Uriel; Morales Sánchez, Israel

    2007-03-01

    The aim of this study was to quantify by immunohistochemistry the number of myoepithelial cells (MyECs) in pleomorphic adenomas (PAs). We retrieved the paraffin cubes of 27 PAs, new slides were done and they were stained with anti-S100 protein antibody. The amount of S-100 protein positive cells was quantified, their morphology was recorded and comparison among MyEC number with age, gender and involved gland were also done. With S-100 protein, MyECs in normal salivary gland tissue were seen surrounding the ductual structures only. In the analysed PAs a mean of 27.4% of the neoplastic cells were positive to the antibody. With the exception of one PA, in all the analysed cases the plasmacytoid cells were the most commonly identified cells (48,6%). Results of this study suggest that MyECs do not constitute the main cellular component of the neoplastic compartment in PAs and corroborate the previously reported evidence by different authors, who studying the PAs suggested that MyECs does not comprise the main cellular neoplastic component of these entities.

  15. In-vacuum sensors for the beamline components of the ITER neutral beam test facility

    Energy Technology Data Exchange (ETDEWEB)

    Dalla Palma, M., E-mail: mauro.dallapalma@igi.cnr.it; Pasqualotto, R.; Spagnolo, S.; Spolaore, M. [Consorzio RFX, Padova 35127 (Italy); Sartori, E. [Consorzio RFX, Padova 35127 (Italy); Università degli Studi di Padova, Padova 35122 (Italy); Veltri, P. [Consorzio RFX, Padova 35127 (Italy); INFN-LNL, Legnaro (PD) 35020 (Italy)

    2016-11-15

    Embedded sensors have been designed for installation on the components of the MITICA beamline, the prototype ITER neutral beam injector (Megavolt ITER Injector and Concept Advancement), to derive characteristics of the particle beam and to monitor the component conditions during operation for protection and thermal control. Along the beamline, the components interacting with the particle beam are the neutralizer, the residual ion dump, and the calorimeter. The design and the positioning of sensors on each component have been developed considering the expected beam-surface interaction including non-ideal and off-normal conditions. The arrangement of the following instrumentation is presented: thermal sensors, strain gages, electrostatic probes including secondary emission detectors, grounding shunt for electrical currents, and accelerometers.

  16. Design evolution and integration of the ITER in-vessel components

    International Nuclear Information System (INIS)

    Martin, A.; Calcagno, B.; Chappuis, Ph.; Daly, E.; Dellopoulos, G.; Furmanek, A.; Gicquel, S.; Heitzenroeder, P.; Jiming, Chen; Kalish, M.; Kim, D.-H.; Khomiakov, S.; Labusov, A.; Loarte, A.; Loughlin, M.; Merola, M.; Mitteau, R.; Polunovski, E.; Raffray, R.; Sadakov, S.

    2013-01-01

    Highlights: ► The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. ► A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. ► The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. ► The blanket manifold system has been redesigned to improve leak detection and localisation. ► The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. -- Abstract: The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. The blanket manifold system has been redesigned to improve leak detection and localisation. The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. This paper describes the status of the redesign of the in-vessel components and the associated integration issues

  17. VDE/disruption EM analysis for ITER in-vessel components

    International Nuclear Information System (INIS)

    Miki, N.; Ioki, K.; Ilio, F.; Kodama, T.; Chiocchio, S.; Williamson, D.; Roccella, M.; Barabaschi, P.; Sayer, R.S.

    1998-01-01

    This paper summarises the results of EM analyses for ITER in-vessel components, such as blanket modules, backplate and divertor modules. In the ITER design the following two disruption scenarios are taken into account: centered or radial disruption, and vertical displacement event (VDE). Eddy currents and forces due to plasma disruption were calculated using the 3D shell element code EDDYCUFF and the 3D solid element code EMAS. The plasma motion and current decay used in the EM analysis was supplied by 2-D axisymmetric plasma equilibrium codes, TSC and MAXFEA. (authors)

  18. Solutions to mitigate heat loads due to electrons on sensitive components of ITER HNB beamlines

    Energy Technology Data Exchange (ETDEWEB)

    Sartori, Emanuele, E-mail: emanuele.sartori@gmail.com [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), C.so Stati Uniti 4, 35127 Padova (Italy); Veltri, Pierluigi; Dalla Palma, Mauro; Agostinetti, Piero [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), C.so Stati Uniti 4, 35127 Padova (Italy); Hemsworth, Ronald; Singh, Mahendrajit [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Serianni, Gianluigi [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), C.so Stati Uniti 4, 35127 Padova (Italy)

    2016-11-01

    Highlights: • Energetic electrons leaking out of the ITER HNB accelerator are simulated. • Electrons generated along the ITER HNB beamline are simulated. • Heat loads and heat load maps on cryopumps are calculated for ITER HNB and test facility. • Protection solutions that will be installed are presented and their effect discussed. - Abstract: The operation of neutral beam injectors for plasma heating and current drive in a fusion device provides challenges in the thermal management of beamline components. Sensitive components such as the cryogenic pumps at beamline periphery shall be protected from the heat flux due to stray electrons. These are emitted by the negative ion accelerator or generated along the beamline by interaction of fast electrons, ions or atoms with background gas and surfaces. In this article the case of the ITER Heating Neutral Beam (HNB) and its test facility MITICA is discussed, for which the beam parameters and the required pulse length of one hour is a major leap forward with respect to the present experience with neutral beam systems. The engineering solutions adopted for effective cryopump protection against the heat load from electrons are described. The use of three-dimensional numerical simulations of particle trajectories in the complex geometry of the beamline was needed for the quantitative estimations of the heat loads. The presented solutions were optimized to minimize the impact on gas pumping and on the functionality of other components.

  19. Development of an original active thermography method adapted to ITER plasma facing components control

    Energy Technology Data Exchange (ETDEWEB)

    Durocher, A.; Vignal, N.; Escourbiac, F.; Farjon, J.L.; Schlosser, J. [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee, 13 - Saint-Paul-lez-Durance (France); Cismondi, F. [Toulon Univ., 83 - La Garde (France)

    2004-07-01

    Among all Non-Destructive Examinations (NDE), active infrared thermography is becoming recognised as a technique available today for improving quality control of many materials and structures involved in heat transfer. The infrared thermography allows to characterise the bond between two materials having different thermal physical properties. In order to increase the defect detection limit of the SATIR test bed, several possibilities have been evaluated to improve the infrared thermography inspection. The implementation in 2003 of a micro-bolometer camera and the improving of the thermo-signal process allowed to increase considerably the detection sensitivity of the SATIR facility. The quality, the spatial stability of infrared image and the detection of edge defect have been also improved. The coupling on the same test bed of SATIR method with a lock-in thermography will be evaluated in this paper. An improvement of the global reliability is expected by data merging produced by the two thermal excitation sources. A new enhanced facility named SATIRPACA has been designed for the full Non Destructive Examination of the High Heat Flux ITER components taking into account these main improvements. These systematic acceptance tests obviously need tools for quality control of critical parts. (authors)

  20. Development of an original active thermography method adapted to ITER plasma facing components control

    International Nuclear Information System (INIS)

    Durocher, A.; Vignal, N.; Escourbiac, F.; Farjon, J.L.; Schlosser, J.; Cismondi, F.

    2004-01-01

    Among all Non-Destructive Examinations (NDE), active infrared thermography is becoming recognised as a technique available today for improving quality control of many materials and structures involved in heat transfer. The infrared thermography allows to characterise the bond between two materials having different thermal physical properties. In order to increase the defect detection limit of the SATIR test bed, several possibilities have been evaluated to improve the infrared thermography inspection. The implementation in 2003 of a micro-bolometer camera and the improving of the thermo-signal process allowed to increase considerably the detection sensitivity of the SATIR facility. The quality, the spatial stability of infrared image and the detection of edge defect have been also improved. The coupling on the same test bed of SATIR method with a lock-in thermography will be evaluated in this paper. An improvement of the global reliability is expected by data merging produced by the two thermal excitation sources. A new enhanced facility named SATIRPACA has been designed for the full Non Destructive Examination of the High Heat Flux ITER components taking into account these main improvements. These systematic acceptance tests obviously need tools for quality control of critical parts. (authors)

  1. Powerloads on the front end components and the duct of the heating and diagnostic neutral beam lines at ITER

    Energy Technology Data Exchange (ETDEWEB)

    Singh, M. J.; Boilson, D.; Hemsworth, R. S.; Geli, F.; Graceffa, J.; Urbani, M.; Schunke, B.; Chareyre, J. [ITER Organisation, 13607 St. Paul-Lez-Durance Cedex (France); Dlougach, E.; Krylov, A. [RRC Kurchatov institute, 1, Kurchatov Sq, Moscow, 123182 (Russian Federation)

    2015-04-08

    The heating and current drive beam lines (HNB) at ITER are expected to deliver ∼16.7 MW power per beam line for H beams at 870 keV and D beams at 1 MeV during the H-He and the DD/DT phases of ITER operation respectively. On the other hand the diagnostic neutral beam (DNB) line shall deliver ∼2 MW power for H beams at 100 keV during both the phases. The path lengths over which the beams from the HNB and DNB beam lines need to be transported are 25.6 m and 20.7 m respectively. The transport of the beams over these path lengths results in beam losses, mainly by the direct interception of the beam with the beam line components and reionisation. The lost power is deposited on the surfaces of the various components of the beam line. In order to ensure the survival of these components over the operational life time of ITER, it is important to determine to the best possible extent the operational power loads and power densities on the various surfaces which are impacted by the beam in one way or the other during its transport. The main factors contributing to these are the divergence of the beamlets and the halo fraction in the beam, the beam aiming, the horizontal and vertical misalignment of the beam, and the gas profile along the beam path, which determines the re-ionisation loss, and the re-ionisation cross sections. The estimations have been made using a combination of the modified version of the Monte Carlo Gas Flow code (MCGF) and the BTR code. The MCGF is used to determine the gas profile in the beam line and takes into account the active gas feed into the ion source and neutraliser, the HNB-DNB cross over, the gas entering the beamline from the ITER machine, the additional gas atoms generated in the beam line due to impacting ions and the pumping speed of the cryopumps. The BTR code has been used to obtain the power loads and the power densities on the various surfaces of the front end components and the duct modules for different scenarios of ITER

  2. Main photoautotrophic components of biofilms in natural draft cooling towers.

    Science.gov (United States)

    Hauer, Tomáš; Čapek, Petr; Böhmová, Petra

    2016-05-01

    While photoautotrophic organisms are an important component of biofilms that live in certain regions of natural draft cooling towers, little is known about these communities. We therefore examined 18 towers at nine sites to identify the general patterns of community assembly in three distinct tower parts, and we examined how community structures differ depending on geography. We also compared the newly acquired data with previously published data. The bottom sections of draft cooling towers are mainly settled by large filamentous algae, primarily Cladophora glomerata. The central portions of towers host a small amount of planktic algae biomass originating in the cooling water. The upper fourths of towers are colonized by biofilms primarily dominated by cyanobacteria, e.g., members of the genera Gloeocapsa and Scytonema. A total of 41 taxa of phototrophic microorganisms were identified. Species composition of the upper fourth of all towers was significantly affected by cardinal position. There was different species composition at positions facing north compared to positions facing south. West- and east-facing positions were transitory and highly similar to each other in terms of species composition. Biofilms contribute to the degradation of paint coatings inside towers.

  3. Modelling ELM heat flux deposition on the ITER main chamber wall

    Czech Academy of Sciences Publication Activity Database

    Kočan, M.; Pitts, R.A.; Lisgo, S.W.; Loarte, A.; Gunn, J. P.; Fuchs, Vladimír

    2015-01-01

    Roč. 463, July (2015), s. 709-713 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI)/21./. Kanazawa, 26.05.2014-30.05.2014] Institutional support: RVO:61389021 Keywords : ELM * ITER Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015

  4. Prediction for disruption erosion of ITER plasma facing components; a comparison of experimental and numerical results

    International Nuclear Information System (INIS)

    Laan, J.G. van der; Akiba, M.; Seki, M.; Hassanein, A.; Tanchuk, V.

    1991-01-01

    An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating paramters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in the thermal quench phase. (orig.)

  5. An operational non destructive examination for ITER divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Durocher, A.; Escourbiac, F.; Farjon, J.L.; Vignal, N.; Cismondi, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [ITER International Team, Cadarache, 13 - St Paul Lez Durance (France); Riccardi, B. [CEFDA CSU-Garching, Garching bei Munchen (Germany)

    2007-07-01

    Full text of publication follows: To meet the power exhaust - heat flux of 20 MW/m{sup 2} - requirements of Plasma Facing Components (PFCs) during plasma operation requires control of their thermal and mechanical integrity. As heat exhaust capability and lifetime of PFCs during in-situ operation are linked to the manufacturing quality, it is an absolute requirement to develop reliable nondestructive examination methods, in particular of the CFC-CuCrZr joint, throughout the manufacturing process. Within the framework of Tokamak Tore Supra upgrade, a pioneering activity has been developed to evaluate the capability of the PFC to be efficiently cooled. In 1998 a test bed - so called SATIR - based on the heat transient method was developed by the CEA and is used today as an inspection tool in order to guarantee the PFCs performances. The technical procurement plan of ITER Divertor targets stated that all Cu cast layers on CFC armour should be subjected to 100% thermographic examination. Each ITER Party should demonstrate its technical capability to carry out the PFC with the required cooling efficiently. The ITER Divertor PFCs pose new challenges especially for the mono-block CFC thickness, and the number of full scale units to be tested which is higher than on any existing or under construction fusion machine. The SATIR method as functional inspection has been identified as the basis test to decide upon the final acceptance of the Divertor PFCs. In order to increase the detection sensitivity of SATIR test bed, several possibilities have been assessed i) the increase of the convective heat transfer coefficient, which improved in a significant way the sensitivity of SATIR diagnostic on ITER components. ii) the installation of a digital infrared camera and the improvement of the thermal signal processing, has led to a considerable increase of performances iii) an innovative process based on spatial image autocorrelation will allow to localize the interlayer defect

  6. Initial results for a 170 GHz high power ITER waveguide component test stand

    Science.gov (United States)

    Bigelow, Timothy; Barker, Alan; Dukes, Carl; Killough, Stephen; Kaufman, Michael; White, John; Bell, Gary; Hanson, Greg; Rasmussen, Dave

    2014-10-01

    A high power microwave test stand is being setup at ORNL to enable prototype testing of 170 GHz cw waveguide components being developed for the ITER ECH system. The ITER ECH system will utilize 63.5 mm diameter evacuated corrugated waveguide and will have 24 >150 m long runs. A 170 GHz 1 MW class gyrotron is being developed by Communications and Power Industries and is nearing completion. A HVDC power supply, water-cooling and control system has been partially tested in preparation for arrival of the gyrotron. The power supply and water-cooling system are being designed to operate for >3600 second pulses to simulate the operating conditions planned for the ITER ECH system. The gyrotron Gaussian beam output has a single mirror for focusing into a 63.5 mm corrugated waveguide in the vertical plane. The output beam and mirror are enclosed in an evacuated duct with absorber for stray radiation. Beam alignment with the waveguide is a critical task so a combination of mirror tilt adjustments and a bellows for offsets will be provided. Analysis of thermal patterns on thin witness plates will provide gyrotron mode purity and waveguide coupling efficiency data. Pre-prototype waveguide components and two dummy loads are available for initial operational testing of the gyrotron. ORNL is managed by UT-Battelle, LLC, for the U.S. Dept. of Energy under Contract DE-AC-05-00OR22725.

  7. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    International Nuclear Information System (INIS)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A; Constans, S; Merola, M; Riccardi, B

    2009-01-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  8. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Science.gov (United States)

    Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

    2009-12-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  9. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A [CEA/IRFM, F-13108, Saint-Paul-lez-Durance (France); Constans, S [AREVA-NP, Le Creusot (France); Merola, M [ITER Organization, Cadarache (France); Riccardi, B [Fusion For Energy, Barcelona (Spain)], E-mail: frederic.escourbiac@cea.fr

    2009-12-15

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  10. Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

    International Nuclear Information System (INIS)

    Bisio, M.; Branca, V.; Marco, M. Di; Federici, A.; Grattarola, M.; Gualco, G.; Guarnone, P.; Luconi, U.; Merola, M.; Ozzano, C.; Pasquale, G.; Poggi, P.; Rizzo, S.; Varone, F.

    2005-01-01

    A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions

  11. Assessment and selection of materials for ITER in-vessel components

    Science.gov (United States)

    Kalinin, G.; Barabash, V.; Cardella, A.; Dietz, J.; Ioki, K.; Matera, R.; Santoro, R. T.; Tivey, R.; ITER Home Teams

    2000-12-01

    During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)-IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti-6Al-4V alloy and two copper alloys, CuCrZr-IG and CuAl25-IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

  12. A review of the US joining technologies for plasma facing components in the ITER fusion reactor

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C.H.; Watson, R.D.; Slattery, K.T.

    1998-02-01

    This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (CuNiBe, OD copper, CuCrZr). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report

  13. Thermal analysis of the in-vessel components of the ITER plasma-position reflectometry

    Energy Technology Data Exchange (ETDEWEB)

    Quental, P. B., E-mail: pquental@ipfn.tecnico.ulisboa.pt; Policarpo, H.; Luís, R.; Varela, P. [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal)

    2016-11-15

    The ITER plasma position reflectometry system measures the edge electron density profile of the plasma, providing real-time supplementary contribution to the magnetic measurements of the plasma-wall distance. Some of the system components will be in direct sight of the plasma and therefore subject to plasma and stray radiation, which may cause excessive temperatures and stresses. In this work, thermal finite element analysis of the antenna and adjacent waveguides is conducted with ANSYS V17 (ANSYS® Academic Research, Release 17.0, 2016). Results allow the identification of critical temperature points, and solutions are proposed to improve the thermal behavior of the system.

  14. Development of the plasma facing components for the dome-liner component of the ITER divertor

    International Nuclear Information System (INIS)

    Luconi, U.; Di Marco, M.; Federici, A.; Grattarola, M.; Gualco, G.; Larrea, J.M.; Merola, M.; Ozzano, C.; Pasquale, G.

    2005-01-01

    On the basis of the design and the specification of the dome-liner elaborated by EFDA, a manufacturing route based on high temperature brazing has been developed and proved by means of the fabrication and testing of several samples and mock-ups. The dome is protected with tungsten armour tiles joined onto heat sinks obtained from a bimetallic plate made of precipitation hardened copper-chromium-zirconium alloy and stainless steel realized by explosion bonding. The brazed joint between the tungsten tiles and the heat sink has been qualified by means of thermal fatigue tests on small-scale mock-ups in reactor relevant conditions. The properties of the explosion bonding joint between the front copper alloy plate to the rear steel backing has been assessed by means of an extensive metallurgical and mechanical test program according to the specification provided by EFDA. The dimensional stability during the fabrication route has been investigated by means of the realization of a relevant curved component that has been dimensionally tested after the completion of each step of the manufacturing route. The results of the experimental activity are presented and discussed in this paper

  15. Millimetre wave attenuation of prototype diagnostic components for the ITER bolometers

    Energy Technology Data Exchange (ETDEWEB)

    Meister, H., E-mail: meister@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Kasparek, W. [Universität Stuttgart, Institut für Grenzflächenverfahrenstechnik & Plasmatechnologie, Stuttgart (Germany); Zhang, D.; Hirsch, M. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Greifswald (Germany); Koll, J. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Zeitler, A. [Universität Stuttgart, Institut für Grenzflächenverfahrenstechnik & Plasmatechnologie, Stuttgart (Germany)

    2015-10-15

    Highlights: • Attenuation of ECRH stray radiation in ITER demonstrated for bolometer prototypes. • Collimator with microwave reflecting grid achieves >70 dB at 170 GHz (ITER frequency). • For frequencies >250 GHz (ECE radiation) ceramic coating showed 40 dB attenuation. • Good shielding at joints of components is prerequisite to prevent microwave leakage. • These methods prevent the impact of ECRH stray radiation on bolometer measurements. - Abstract: Bolometers in current and future fusion devices, in particular those in ITER, are vulnerable to stray radiation from electron cyclotron resonance heating (ECRH) which results in measurement errors for plasma radiation detection. To protect the detectors from this stray radiation in the millimetre wavelength range, dedicated diagnostic components have been designed and tested. One option is to place a top plate which contains a microwave-reflecting grid onto the collimators. Another option investigated is the coating of the collimator channels using a microwave absorbing ceramic. Measurements of the mm-wave attenuation of the collimator in front of the bolometer detectors with and without top plate or coated collimator channels have been performed in the frequency range of 125–420 GHz. The attenuation factor of the collimator channels at 170 GHz (the ECRH frequency for ITER) with neither microwave grid nor coating is typically 10 dB. The coating enhances this to 40 dB and including the microwave grid yields at least an attenuation factor of 70 dB, which is sufficient to reduce the residual ECRH induced signal significantly below the one due to plasma radiation. Placing a bolometer camera (collimator connected to detector housing) inside the isotropic microwave field of the test facility MISTRAL, the attenuation factor of the full diagnostic set-up using a top plate was determined to be in the order of 45 dB. This degraded attenuation implies that particular attention has to be paid to design and quality

  16. Specification of properties and design allowables for copper alloys used in HHF components of ITER

    DEFF Research Database (Denmark)

    Kalinin, G.M.; Fabritziev, S.A.; Singh, B.N.

    2002-01-01

    CrZr and CuAl25 are not yet fully characterised. The performed R&D gives a basis for the specification of physical and mechanical properties required for the design analysis in accordance with the ITER Structural Design Criteria for In-vessel Components (SDC-IC). For both CuCrZr-IG and CuAl25-IG alloys......Two types of copper alloys, precipitation hardened (PH) Cu (CuCrZr-IG) and dispersion strengthened (DS) Cu (CuAl25-IG), are proposed as heat sink materials for the high heat flux (HHF) components of ITER. However, copper alloys are not included in any national codes, and properties of both Cu......, the statistical evaluation of available experimental data has been used to calculate the temperature dependence of the average value and of the 95% confidence limit of tensile properties. The stress limits, Sm, Se, and Sd, have been estimated on the basis of available data. The procedure used for specification...

  17. Procurement of the ITER magnet components supplied by F4E

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  18. 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.

  19. 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.

  20. Progress on corrugated waveguide components suitable for ITER ECH&CD transmission lines

    Directory of Open Access Journals (Sweden)

    Murphy C.J.

    2012-09-01

    Full Text Available The 2 MW cw requirement for ITER EC transmission lines has led GA to design a number of new and modified components, namely power monitors and polarizers to meet this requirement. In addition, the ITER transmission lines may require sliding joints or bellows to accommodate the thermal expansion of the lines and/or vessel motion. The ex-vessel sections of the EC launchers require double seal waveguides, miter bends and possibly double seal sliding joints to assure tritium retention in this region. GA has developed designs for many of these components and some of them have already been tested at high power at the JAEA RF Test Stand. Thermal analyses of the standard GA power monitor miter bend show that our standard design is suitable for 1 MW cw operation when the H-field is in the plane of the miter bend. For 2 MW cw operation a modified design, which also requires H-plane orientation, has been developed and a prototype has been fabricated and is ready for testing. For long pulse/cw operation, a 2 MW calorimetric miter bend with thermally isolated mirror has been designed and a prototype has been fabricated. Since the mirror is thermally isolated, calorimetry on the mirror cooling water can provide a measure of absorbed power. Such a miter bend, when made in a double seal version, could be useful for monitoring total power at the end of an EC transmission line just before the in-vessel section of an EC launcher. A mode analyzer using an all metal water-cooled beam splitter is being developed for use in measuring in real time the HE11 and higher order mode content or total power in EC transmission lines. Such a high power diagnostic can be very useful in optimizing the alignment of the MOU output into a transmission line and in monitoring the HE11 mode purity at various locations in a transmission line. Prior to making a 63.5 mm 170 GHz version suitable for use on ITER, several prototype 31.75 mm 110 GHz versions have been fabricated and tested at

  1. The Research of Multiple Attenuation Based on Feedback Iteration and Independent Component Analysis

    Science.gov (United States)

    Xu, X.; Tong, S.; Wang, L.

    2017-12-01

    How to solve the problem of multiple suppression is a difficult problem in seismic data processing. The traditional technology for multiple attenuation is based on the principle of the minimum output energy of the seismic signal, this criterion is based on the second order statistics, and it can't achieve the multiple attenuation when the primaries and multiples are non-orthogonal. In order to solve the above problems, we combine the feedback iteration method based on the wave equation and the improved independent component analysis (ICA) based on high order statistics to suppress the multiple waves. We first use iterative feedback method to predict the free surface multiples of each order. Then, in order to predict multiples from real multiple in amplitude and phase, we design an expanded pseudo multi-channel matching filtering method to get a more accurate matching multiple result. Finally, we present the improved fast ICA algorithm which is based on the maximum non-Gauss criterion of output signal to the matching multiples and get better separation results of the primaries and the multiples. The advantage of our method is that we don't need any priori information to the prediction of the multiples, and can have a better separation result. The method has been applied to several synthetic data generated by finite-difference model technique and the Sigsbee2B model multiple data, the primaries and multiples are non-orthogonal in these models. The experiments show that after three to four iterations, we can get the perfect multiple results. Using our matching method and Fast ICA adaptive multiple subtraction, we can not only effectively preserve the effective wave energy in seismic records, but also can effectively suppress the free surface multiples, especially the multiples related to the middle and deep areas.

  2. A simple iterative independent component analysis algorithm for vibration source signal identification of complex structures

    Directory of Open Access Journals (Sweden)

    Dong-Sup Lee

    2015-01-01

    Full Text Available Independent Component Analysis (ICA, one of the blind source separation methods, can be applied for extracting unknown source signals only from received signals. This is accomplished by finding statistical independence of signal mixtures and has been successfully applied to myriad fields such as medical science, image processing, and numerous others. Nevertheless, there are inherent problems that have been reported when using this technique: insta- bility and invalid ordering of separated signals, particularly when using a conventional ICA technique in vibratory source signal identification of complex structures. In this study, a simple iterative algorithm of the conventional ICA has been proposed to mitigate these problems. The proposed method to extract more stable source signals having valid order includes an iterative and reordering process of extracted mixing matrix to reconstruct finally converged source signals, referring to the magnitudes of correlation coefficients between the intermediately separated signals and the signals measured on or nearby sources. In order to review the problems of the conventional ICA technique and to vali- date the proposed method, numerical analyses have been carried out for a virtual response model and a 30 m class submarine model. Moreover, in order to investigate applicability of the proposed method to real problem of complex structure, an experiment has been carried out for a scaled submarine mockup. The results show that the proposed method could resolve the inherent problems of a conventional ICA technique.

  3. Microstructure investigation of bronze/steel brazed joints proposed for HHF components of ITER manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G.M. [Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)], E-mail: gmk@nikiet.ru; Krestnikov, N.S. [Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Jarovinskiy, Yu.L.; Makhin, I.D.; Nikolaev, V.V. [Rocket and Space Corporation ' Energia' , Korolev, Lenina Street 4-a, Moscow 141070 (Russian Federation); Skladnov, K.S.; Strebkov, Yu.S.; Zolotarev, V.B. [Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)

    2008-12-15

    Brazing is considered as one of the perspective option of high heat flux components of ITER manufacturing. CuCrZr bronze, austenitic steel AISI 321-type and PM-17-type (Ni-Mn-Fe-Si-Sn-B alloy) brazed material were used for the development of brazing technology. Two type of brazing have been studied within the framework of recent investigation: - Hot isostatic pressing (HIP)-assisted brazing. - Furnace-assisted brazing (with uniaxial compression loading). For the hydrostatic pressing (HIP) the brazed components were pressed out for about 175 MPa during 2.5 h at the temperature 1035-1040 deg. C. For the furnace-assisted brazing all components were inserted into the sealed can, vacuumed and heated up to brazing temperature {approx}950 deg. C. Fast cooling and ageing heat treatment (500 deg. C and 4 h) were applied to provide high strength of CuCrZr bronze. Microsections of specimens cut from the joints were studied by optical microscopy and by scanning electron microscopy (SEM). The microstructure, distribution of alloying elements of base metals and of brazed material components were studied in the joints. Results of these studies are discussed in this paper. The data shows that there is a potential for using more simple and cheap (in comparison with common HIP) technologies of bronze to steel joining with satisfactory quality.

  4. Development of radiation hard components for ITER blanket remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Makiko, E-mail: saito.makiko@jaea.go.jp; Anzai, Katsunori; Maruyama, Takahito; Noguchi, Yuto; Ueno, Kenichi; Takeda, Nobukazu; Kakudate, Satoshi

    2016-11-01

    Highlights: • Clarify the components that will degrade by gamma ray irradiation. • Perform the irradiation tests to BRHS components. • Optimize the materials to increase the radiation hardness. - Abstract: The ITER blanket remote handling system (BRHS) will be operated in a high radiation environment (250 Gy/h max.) and must stably handle the blanket modules, which weigh 4.5 t and are more than 1.5 m in length, with a high degree of position and posture accuracy. The reliability of the system can be improved by reviewing the failure events of the system caused by high radiation. A failure mode and effects analysis (FMEA) identified failure modes and determined that lubricants, O-rings, and electric insulation cables were the dominant components affecting radiation hardness. Accordingly, we tried to optimize the lubricants and cables of the AC servo motors by using polyphenyl ether (PPE)-based grease and polyether ether ketone (PEEK), respectively. Materials containing radiation protective agents were also selected for the cable sheaths and O-rings to improve radiation hardness. Gamma ray irradiation tests were performed on these components and as a result, a radiation hardness of 8 MGy was achieved for the AC servo motors. On the other hand, to develop the radiation hardness and BRHS compatibility furthermore, the improvement of materials of cable and O ring were performed.

  5. Overview of the EU small scale mock-up tests for ITER high heat flux components

    International Nuclear Information System (INIS)

    Vieider, G.; Barabash, V.; Cardella, A.

    1998-01-01

    This task within the EU R and D for ITER was aimed at the development of basic manufacturing solutions for the high heat flux plasma facing components such as the divertor targets, the baffles and limiters. More than 50 representative small-scale mock-ups have been manufactured with beryllium, carbon and tungsten armour using various joining technologies. High heat flux testing of 20 of these mock-ups showed the carbon mono-blocks to be the most robust solution, surviving 2000 cycles at absorbed heat fluxes of up to 24 MW m -2 . With flat armour tiles rapid joint failures occurred at 5-16 MW m -2 depending on joining technology and armour material. These test results serve as a basis for the selection of manufacturing options and materials for the prototypes now being ordered. (orig.)

  6. Transmission line component testing for the ITER Ion Cyclotron Heating and Current Drive System

    Science.gov (United States)

    Goulding, Richard; Bell, G. L.; Deibele, C. E.; McCarthy, M. P.; Rasmussen, D. A.; Swain, D. W.; Barber, G. C.; Barbier, C. N.; Cambell, I. H.; Moon, R. L.; Pesavento, P. V.; Fredd, E.; Greenough, N.; Kung, C.

    2014-10-01

    High power RF testing is underway to evaluate transmission line components for the ITER Ion Cyclotron Heating and Current Drive System. The transmission line has a characteristic impedance Z0 = 50 Ω and a nominal outer diameter of 305 mm. It is specified to carry up to 6 MW at VSWR = 1.5 for 3600 s pulses, with transient voltages up to 40 kV. The transmission line is actively cooled, with turbulent gas flow (N2) used to transfer heat from the inner to outer conductor, which is water cooled. High voltage and high current testing of components has been performed using resonant lines generating steady state voltages of 35 kV and transient voltages up to 60 kV. A resonant ring, which has operated with circulating power of 6 MW for 1 hr pulses, is being used to test high power, low VSWR operation. Components tested to date include gas barriers, straight sections of various lengths, and 90 degree elbows. Designs tested include gas barriers fabricated from quartz and aluminum nitride, and transmission lines with quartz and alumina inner conductor supports. The latest results will be presented. This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy.

  7. Comprehensive simulation of vertical plasma instability events and their serious damage to ITER plasma facing components

    International Nuclear Information System (INIS)

    Hassanein, A.; Sizyuk, T.

    2008-01-01

    Safe and reliable operation is still one of the major challenges in the development of the new generation of ITER-like fusion reactors. The deposited plasma energy during major disruptions, edge-localized modes (ELMs) and vertical displacement events (VDEs) causes significant surface erosion, possible structural failure and frequent plasma contamination. While plasma disruptions and ELM will have no significant thermal effects on the structural materials or coolant channels because of their short deposition time, VDEs having longer-duration time could have a destructive impact on these components. Therefore, modelling the response of structural materials to VDE has to integrate detailed energy deposition processes, surface vaporization, phase change and melting, heat conduction to coolant channels and critical heat flux criteria at the coolant channels. The HEIGHTS 3D upgraded computer package considers all the above processes to specifically study VDE in detail. Results of benchmarking with several known laboratory experiments prove the validity of HEIGHTS implemented models. Beryllium and tungsten are both considered surface coating materials along with copper structure and coolant channels using both smooth tubes with swirl tape insert. The design requirements and implications of plasma facing components are discussed along with recommendations to mitigate and reduce the effects of plasma instabilities on reactor components.

  8. Investigation on bonding defects in ITER first wall beryllium armour components by combining analytical and experimental methods

    Energy Technology Data Exchange (ETDEWEB)

    Pérez, Germán, E-mail: german.perez.pichel@gmail.com; Mitteau, Raphaël; Eaton, Russell; Raffray, René

    2015-12-15

    Highlights: • Bonding defects at the ITER first wall beryllium armour are studied. • Experimental and analytical methods are combined. • Models supporting test results interpretation are proposed. • Guidelines for new experimental protocols are suggested. • Contribution to the definition of defects acceptance criteria. - Abstract: The reliability of the plasma facing components (PFCs) is essential for the efficient plasma operation in a fusion machine. This concerns especially the bond between the armour tiles facing the plasma and the heat sink material (copper alloy). The different thermal expansions of the bonded materials cause a stress distribution in the bond, which peaks at the bond edge. Under cyclic heat flux and accounting for the possible presence of bonding defects, this stress could reach a level where the component might be jeopardised. Because of the complexity of describing realistically by analyses and models the stress evolution in the bond, “design by experiments” is the main procedure for defining and qualifying the armour joint. Most of the existing plasma operation know-how on actively cooled PFCs has been obtained with carbon composite armour tiles. In ITER, the tiles of the first wall are made out of beryllium, which means that the know-how is progressively adapted to this specific bimetallic pair. Nonetheless, analyses are still performed for supporting the R&D experimental programme. This paper: explores methods for combining experimental results with finite element and statistical analyses; benchmarks test results; proposes hypothesis and rationales consistent with test results interpretations; suggests guidelines for defining possible further experimental protocols; and contributes to the definition of defects acceptance criteria.

  9. Investigation on bonding defects in ITER first wall beryllium armour components by combining analytical and experimental methods

    International Nuclear Information System (INIS)

    Pérez, Germán; Mitteau, Raphaël; Eaton, Russell; Raffray, René

    2015-01-01

    Highlights: • Bonding defects at the ITER first wall beryllium armour are studied. • Experimental and analytical methods are combined. • Models supporting test results interpretation are proposed. • Guidelines for new experimental protocols are suggested. • Contribution to the definition of defects acceptance criteria. - Abstract: The reliability of the plasma facing components (PFCs) is essential for the efficient plasma operation in a fusion machine. This concerns especially the bond between the armour tiles facing the plasma and the heat sink material (copper alloy). The different thermal expansions of the bonded materials cause a stress distribution in the bond, which peaks at the bond edge. Under cyclic heat flux and accounting for the possible presence of bonding defects, this stress could reach a level where the component might be jeopardised. Because of the complexity of describing realistically by analyses and models the stress evolution in the bond, “design by experiments” is the main procedure for defining and qualifying the armour joint. Most of the existing plasma operation know-how on actively cooled PFCs has been obtained with carbon composite armour tiles. In ITER, the tiles of the first wall are made out of beryllium, which means that the know-how is progressively adapted to this specific bimetallic pair. Nonetheless, analyses are still performed for supporting the R&D experimental programme. This paper: explores methods for combining experimental results with finite element and statistical analyses; benchmarks test results; proposes hypothesis and rationales consistent with test results interpretations; suggests guidelines for defining possible further experimental protocols; and contributes to the definition of defects acceptance criteria.

  10. Automated main-chain model building by template matching and iterative fragment extension

    International Nuclear Information System (INIS)

    Terwilliger, Thomas C.

    2003-01-01

    A method for automated macromolecular main-chain model building is described. An algorithm for the automated macromolecular model building of polypeptide backbones is described. The procedure is hierarchical. In the initial stages, many overlapping polypeptide fragments are built. In subsequent stages, the fragments are extended and then connected. Identification of the locations of helical and β-strand regions is carried out by FFT-based template matching. Fragment libraries of helices and β-strands from refined protein structures are then positioned at the potential locations of helices and strands and the longest segments that fit the electron-density map are chosen. The helices and strands are then extended using fragment libraries consisting of sequences three amino acids long derived from refined protein structures. The resulting segments of polypeptide chain are then connected by choosing those which overlap at two or more C α positions. The fully automated procedure has been implemented in RESOLVE and is capable of model building at resolutions as low as 3.5 Å. The algorithm is useful for building a preliminary main-chain model that can serve as a basis for refinement and side-chain addition

  11. Automated main-chain model building by template matching and iterative fragment extension.

    Science.gov (United States)

    Terwilliger, Thomas C

    2003-01-01

    An algorithm for the automated macromolecular model building of polypeptide backbones is described. The procedure is hierarchical. In the initial stages, many overlapping polypeptide fragments are built. In subsequent stages, the fragments are extended and then connected. Identification of the locations of helical and beta-strand regions is carried out by FFT-based template matching. Fragment libraries of helices and beta-strands from refined protein structures are then positioned at the potential locations of helices and strands and the longest segments that fit the electron-density map are chosen. The helices and strands are then extended using fragment libraries consisting of sequences three amino acids long derived from refined protein structures. The resulting segments of polypeptide chain are then connected by choosing those which overlap at two or more C(alpha) positions. The fully automated procedure has been implemented in RESOLVE and is capable of model building at resolutions as low as 3.5 A. The algorithm is useful for building a preliminary main-chain model that can serve as a basis for refinement and side-chain addition.

  12. Mock-up test on key components of ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Matsumoto, Yasuhiro; Taguchi, Koh; Kozaka, Hiroshi; Shibanuma, Kiyoshi; Tesini, Alessandro

    2009-01-01

    The maintenance operation of the ITER in-vessel component, such as a blanket and divertor, must be executed by the remote equipment because of the high gamma-ray environment. During the Engineering Design Activity (EDA), the Japan Atomic Energy Agency (then called as Japan Atomic Energy Research Institute) had been fabricated the prototype of the vehicle manipulator system for the blanket remote handling and confirmed feasibility of this system including automatic positioning of the blanket and rail deployment procedure of the articulated rail. The ITER agreement, which entered into force in the last year, formally decided that Japan will procure the blanket remote handling system and the JAEA, as the Japanese Domestic Agency, is continuing several R and Ds so that the system can be procured smoothly. The residual key issues after the EDA are rail connection and cable handling. The mock-ups of the rail connection mechanism and the cable handling system were fabricated from the last year and installed at the JAEA Naka Site in this March. The former was composed of the rail connecting mechanism, two rail segments and their handling systems. The latter one utilized a slip ring, which implemented 80 lines for power and 208 lines for signal, because there is an electrical contact between the rotating spool and the fixed base. The basic function of these systems was confirmed through the mock-up test. The rail connection mechanism, for example, could accept misalignment of 1.5-2 mm at least. The future test plan is also mentioned in the paper.

  13. Ultrasonic techniques for quality assessment of ITER Divertor plasma facing component

    International Nuclear Information System (INIS)

    Martinez-Ona, Rafael; Garcia, Monica; Medrano, Mercedes

    2009-01-01

    The divertor is one of the most challenging components of ITER machine. Its plasma facing components contain thousands of joints that should be assessed to demonstrate their integrity during the required lifetime. Ultrasonic (US) techniques have been developed to study the capability of defect detection and to control the quality and degradation of these interfaces after the manufacturing process. Three types of joints made of carbon fibre composite to copper alloy, tungsten to copper alloy, and copper-to-copper alloy with two types of configurations have been studied. More than 100 samples representing these configurations and containing implanted flaws of different sizes have been examined. US techniques developed are detailed and results of validation samples examination before and after high heat flux (HHF) tests are presented. The results show that for W monoblocks the US technique is able to detect, locate and size the degradations in the two sample joints; for CFC monoblocks, the US technique is also able to detect, locate and size the calibrated defects in the two joints before the HHF, however after the HHF test the technique is not able to reliably detect defects in the CFC/Cu joint; finally, for the W flat tiles the US technique is able to detect, locate and size the calibrated defects in the two joints before HHF test, nevertheless defect location and sizing are more difficult after the HHF test.

  14. Design and R&D for manufacturing the beamline components of MITICA and ITER HNBs

    Energy Technology Data Exchange (ETDEWEB)

    Dalla Palma, M., E-mail: mauro.dallapalma@igi.cnr.it [Consorzio RFX, Padova (Italy); Sartori, E. [Consorzio RFX, Padova (Italy); Blatchford, P.; Chuilon, B. [CCFE, Culham Science Centre, Oxfordshire (United Kingdom); Graceffa, J. [ITER Organization, St Paul Lez Durance (France); Hanke, S. [KIT, Institute for Technical Physics, Eggenstein-Leopoldshafen (Germany); Hardie, C. [CCFE, Culham Science Centre, Oxfordshire (United Kingdom); Masiello, A. [F4E, Barcelona (Spain); Muraro, A. [Consorzio RFX, Padova (Italy); Ochoa, S. [KIT, Institute for Technical Physics, Eggenstein-Leopoldshafen (Germany); Shah, D. [ITER Organization, St Paul Lez Durance (France); Veltri, P.; Zaccaria, P.; Zaupa, M. [Consorzio RFX, Padova (Italy)

    2015-10-15

    Highlights: • Particle beam-component interaction was analysed developing and applying numerical codes. • Gas density distribution was calculated with AVOCADO code and applied for electrical analyses. • High heat flux components were designed, analysed with subcooled boiling, verified for fatigue. • Fracture behaviour of ceramics was analysed by finite element modelling and was verified. • R&D supports the design of the beamline components, especially for water-vacuum barriers. - Abstract: The design of the beamline components of MITICA, the full prototype of the ITER heating neutral beam injectors, is almost finalised and technical specifications for the procurement are under preparation. These components are the gas neutraliser, the electrostatic residual ion dump, and the calorimeter. Electron dump panels are foreseen each side of the upstream end of the neutraliser to protect the cryo-panels from electrons, created by stripping and other processes, that exit the 1 MeV accelerator. As the design of the components must fulfil requirements on the beam physics, insight on physical processes is required to identify performance trade-offs and constraints. The spatial gas distribution was simulated to verify the pumping requirements with electron dump panels and local conditions for breakdown voltage. Electrostatic analyses were carried out for the insulating elements of the RID to verify the limits of the electric field intensity. Different criteria were approached to investigate the fracture behaviour of ceramics considering the manufacturing implications and extrapolating the conditions for proof testing. Severe heating conditions will be applied steadily, as the maximum pulse duration is 1 h, and cyclically so requiring to fulfil fatigue and ratcheting verifications. High heat fluxes, up to 13 MW/m{sup 2} on the calorimeter, with enhanced heat transfer in subcooled boiling conditions will occur in the actively cooled CuCr1Zr panel elements provided with

  15. Modelling and computer simulation for the manufacture by powder HIPing of Blanket Shield components for ITER

    International Nuclear Information System (INIS)

    Gillia, O.; Bucci, Ph.; Vidotto, F.; Leibold, J.-M.; Boireau, B.; Boudot, C.; Cottin, A.; Lorenzetto, P.; Jacquinot, F.

    2006-01-01

    In components of blanket modules for ITER, intricate cooling networks are needed in order to evacuate all heat coming from the plasma. Hot Isostatic Pressing (HIPing) technology is a very convenient method to produce near net shape components with complex cooling network through massive stainless steel parts by bonding together tubes inserted in grooves machined in bulk stainless steel. Powder is often included in the process so as to release difficulties arising with gaps closure between tube and solid part or between several solid parts. In the mean time, it releases the machining precision needed on the parts to assemble before HIP. However, inserting powder in the assembly means densification, i.e. volume change of powder during the HIP cycle. This leads to global and local shape changes of HIPed parts. In order to control the deformations, modelling and computer simulation are used. This modelling and computer simulation work has been done in support to the fabrication of a shield prototype for the ITER blanket. Problems such as global bending of the whole part and deformations of tubes in their powder bed are addressed. It is important that the part does not bend too much. It is important as well to have circular tube shape after HIP, firstly in order to avoid their rupture during HIP but also because non destructive ultrasonic examination is needed to check the quality of the densification and bonding between tube and powder or solid parts; the insertions of a probe in the tubes requires a minimal circular tube shape. For simulation purposes, the behaviour of the different materials has to be modelled. Although the modelling of the massive stainless steel behaviour is not neglected, the most critical modelling is about power. For this study, a thorough investigation on the powder behaviour has been performed with some in-situ HIP dilatometry experiments and some interrupted HIP cycles on trial parts. These experiments have allowed the identification of a

  16. Iterative development of Stand Up Australia: a multi-component intervention to reduce workplace sitting

    Science.gov (United States)

    2014-01-01

    Background Sitting, particularly in prolonged, unbroken bouts, is widespread within the office workplace, yet few interventions have addressed this newly-identified health risk behaviour. This paper describes the iterative development process and resulting intervention procedures for the Stand Up Australia research program focusing on a multi-component workplace intervention to reduce sitting time. Methods The development of Stand Up Australia followed three phases. 1) Conceptualisation: Stand Up Australia was based on social cognitive theory and social ecological model components. These were operationalised via a taxonomy of intervention strategies and designed to target multiple levels of influence including: organisational structures (e.g. via management consultation), the physical work environment (via provision of height-adjustable workstations), and individual employees (e.g. via face-to-face coaching). 2) Formative research: Intervention components were separately tested for their feasibility and acceptability. 3) Pilot studies: Stand Up Comcare tested the integrated intervention elements in a controlled pilot study examining efficacy, feasibility and acceptability. Stand Up UQ examined the additional value of the organisational- and individual-level components over height-adjustable workstations only in a three-arm controlled trial. In both pilot studies, office workers’ sitting time was measured objectively using activPAL3 devices and the intervention was refined based on qualitative feedback from managers and employees. Results Results and feedback from participants and managers involved in the intervention development phases suggest high efficacy, acceptance, and feasibility of all intervention components. The final version of the Stand Up Australia intervention includes strategies at the organisational (senior management consultation, representatives consultation workshop, team champions, staff information and brainstorming session with information

  17. Energy deposition and thermal effects of runaway electrons in ITER-FEAT plasma facing components

    International Nuclear Information System (INIS)

    Maddaluno, G.; Maruccia, G.; Merola, M.; Rollet, S.

    2003-01-01

    The profile of energy deposited by runaway electrons (RAEs) of 10 or 50 MeV in International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) plasma facing components (PFCs) and the subsequent temperature pattern have been calculated by using the Monte Carlo code FLUKA and the finite element heat conduction code ANSYS. The RAE energy deposition density was assumed to be 50 MJ/m 2 and both 10 and 100 ms deposition times were considered. Five different configurations of PFCs were investigated: primary first wall armoured with Be, with and without protecting CFC poloidal limiters, both port limiter first wall options (Be flat tile and CFC monoblock), divertor baffle first wall, armoured with W. The analysis has outlined that for all the configurations but one (port limiter with Be flat tile) the heat sink and the cooling tube beneath the armour are well protected for both RAE energies and for both energy deposition times. On the other hand large melting (W, Be) or sublimation (C) of the surface layer occurs, eventually affecting the PFCs lifetime

  18. Energy deposition and thermal effects of runaway electrons in ITER-FEAT plasma facing components

    Science.gov (United States)

    Maddaluno, G.; Maruccia, G.; Merola, M.; Rollet, S.

    2003-03-01

    The profile of energy deposited by runaway electrons (RAEs) of 10 or 50 MeV in International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) plasma facing components (PFCs) and the subsequent temperature pattern have been calculated by using the Monte Carlo code FLUKA and the finite element heat conduction code ANSYS. The RAE energy deposition density was assumed to be 50 MJ/m 2 and both 10 and 100 ms deposition times were considered. Five different configurations of PFCs were investigated: primary first wall armoured with Be, with and without protecting CFC poloidal limiters, both port limiter first wall options (Be flat tile and CFC monoblock), divertor baffle first wall, armoured with W. The analysis has outlined that for all the configurations but one (port limiter with Be flat tile) the heat sink and the cooling tube beneath the armour are well protected for both RAE energies and for both energy deposition times. On the other hand large melting (W, Be) or sublimation (C) of the surface layer occurs, eventually affecting the PFCs lifetime.

  19. The ITER Management Advisory Committee (MAC) meeting in Garching

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    1999-01-01

    The ITER management advisory committee meeting was held on 22-23 July 1999 in Garching, Germany. The main topics were the ITER EDA status, task status summary and work program, joint fund, information technology needs at the ITER joint work sites, the disposition of R and D components and a schedule of ITER meetings

  20. ITER Port Plug Engineering Trainee Program: Design, manufacturing and integration of structural components (analysis of the attachment)

    International Nuclear Information System (INIS)

    Zeile, Christian; Neuberger, Heiko; Dolensky, Bernhard

    2011-01-01

    The structural connection of helium cooled plasma facing components in ITER to the water cooled structural port plug shield requires an attachment system, which is able to cope with two main contradicting requirements: The attachment system has to be rigid in order to withstand mechanical loads, e.g. due to the deadweight or static and transient electro-magnetic loads. On the other hand, the attachment system has to be flexible in order to compensate the different thermal strains in between the plasma facing test devices (300-550 deg. C) and the port plug structure (∼120 deg. C). The paper presents the latest developments of an attachment system consisting of flexible attachment blocks with lamellae. The optimization steps as well as the connection to the shield are described. The results of the thermo-mechanical analyses under a defined worst-case scenario confirm the feasibility of the lamella design. This work has been performed in the frame of the EFDA goal orientated trainee program on port plug engineering.

  1. Development of ITER PRM and standard parts catalogues in CATIA V5 for tritium-containing systems and components

    International Nuclear Information System (INIS)

    Lazar, Alin; Brad, Sebastian; Zamfirache, Marius; Soare, Sorin; Sofalca, Nicolae; Vijulie, Mihai

    2006-01-01

    CATIA V5 is a software chosen to perform the design and integration within ITER of both systems: fluid and mechanical systems. The broad range of applications provides the ability to develop the design process from the functional 2D design (P and ID) to the 3D plant layouts and detailed design. The 2D symbols for the equipment and piping components were developed in accordance with EN ISO 10628 standard, ISO 3511 (part I, II and IV) standard and DIN 28401 standard and considering the classification and settings done by FZK-team in the PRM-FZK-TLK in order to ensure compliance with the established design standards. These symbols were inserted in 2D catalogues linked with the main catalogue installed in PRM-FZK-TLK on CATIA SERVER at TLK-FZK and intensively tested in Piping and Instrumentation Diagrams. The 3D part for the equipment, instruments and piping components was developed according to the specifications, industrial conventions, terminology and literature available in library, internet and practice. All parts were parametrically built in order to allow the designer to modify the part geometry according to the design. The 3D parts were typed in accordance with IKARUS-Project Instructions by FZK team, inserted in 3D test catalogues and tested in Equipment and Systems workbench. The elements of P and ID, pipes, piping parts, instruments and equipment carry attribute information such as fluid type, pressure rating, insulation or material. The list of attributes is defined in the PRM and consists of the standard industrial attributes implemented by default in CATIA V5 and project specific attributes defined by Project Administration. 2D logical design and 3D Piping models are logically related through the common functions definition in the PRM. The task, TW5-TTFD-TPI-51, was performed collaboratively by MEdC/ICIT and FZK/TLK. (authors)

  2. Mechanical and thermal resistance of multi-material components for ITER

    International Nuclear Information System (INIS)

    Burlet, H.

    2013-01-01

    The First Wall panels for ITER are complex parts composed of stainless steel, copper and beryllium [1]. These materials are joined using diffusion bonding technique. The stainless steel is a commonly used in nuclear reactors 316LN material and acts as a structural material. The copper alloy is a CuCrZr material which acts as a heat sink. The beryllium consisting in tiles and layer is used as the protective plasma facing material. The fabrication of these panels is performed through 2 main steps. The first step consists in welding all together a bi-metallic support structure made from a thick CuCrZr plate embedded with 316LN tubes and bonded to a thick 316LN backing plate with cooling channels. The bonding is performed in a HIP (Hot Isostatic Pressure) facility. The second step is performed at a lower temperature and aims at simultaneously welding by HIP Be onto CuCrZr and ageing the CuCrZr heat sink to obtain the correct mechanical resistance of this alloy reinforced by precipitates. The various joints 316LN/316LN, 316LN/CuCrZr, and CuCrZr/Be are then characterized [2] from a microstructural point of view and by mechanical tests. It is quite hard to characterize the strength of a diffusion bonded joints. Standard tests may be used for homogeneous joints whereas specific tests have been developed to characterize the heterogeneous bonds. To optimize the bond, we performed mainly impact and tensile bi-material tests (Fig 1). Once the manufacture parameters have been optimized, advanced mechanical tests are performed based on Bimetallic CT specimens, axisymmetric notched specimens, 4P bending specimens. Numerical simulations are required to analyse the mechanical response. In order to characterize the fatigue resistance of the joints, specific mock-ups have been designed by the European Fusion Development Agreement EFDA team (Fig 2). Results of heat flux testing will be reviewed for the various joints. The assembly of heterogeneous materials by Hipping is very complex

  3. Development of technology and properties investigation of steel/bronze joints proposed for ITER HHF components manufacturing

    International Nuclear Information System (INIS)

    Kalinin, G.; Krestnikov, N.S.; Strebkov, Y.S.; Abramov, V.Y.; Gervash, A.; Mazul, I.; Zolotarev, V.B.; Fabritsiev, S.A.

    2007-01-01

    Full text of publication follows: During the development of ITER HHF components manufacturing it is necessary to provide reliable joints between heat sink material made of CuCrZr bronze and the supporting construction made of austenitic steel. Four different methods have been tried out: - Hot Isostatic Pressing (HIP), - HIP assisted brazing, - furnace assisted brazing, - Casting. The investigation of structure and properties of joints show that HIP and casting provide the better results than the other technologies. However, HIP is relatively expensive technology, and big size HIP furnace is required for the full scale components manufacturing that are not available n RF now. Therefore, casting was selected as a reference manufacturing technology for the primary wall of ITER modules n RF. The paper summarizes the results of bronze/steel joints manufacturing and investigation of their properties. (authors)

  4. Maintenance implications of critical components in ITER CXRS upper port plug design

    NARCIS (Netherlands)

    Koning, J.; Jaspers, R.; Doornink, J.; Ouwehand, B.; Klinkhamer, J.F.F.; Snijders, B.; Sadakov, S.; Heemskerk, C.

    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

  5. Evaluation of material integrity on electricity generator water steam cycles component (Main Steam Pipe)

    International Nuclear Information System (INIS)

    Sudardjo; Histori; Triyadi, Ari

    1998-01-01

    The evaluation of material integrity on electricity generator component has been done. That component was main steam pipe of Unit II Suralaya Coal Fired Power Plant. evaluation was done by replication technique. The damage was found are two porosity's, from two point samples of six points sample population. Based on cavity evaluation in steels, which proposed by Neubauer and Wedel that porosity's still at class A damage. For class A damage, its means no remedial action would be required until next major scheduled maintenance outage. That porosity's was grouped on isolated cavities and not need ti repair that main steam pipe component less than three year after replication test

  6. The Plasma-Facing Components Transporter (PFCT) : a Prototype System for PFC Replacement on the new ITER 2001 Cassette Mock-up

    International Nuclear Information System (INIS)

    Micciche, G.; Lorenzelli, L.; Muro, L.; Irving, M.

    2006-01-01

    The remote maintainability of the early ITER divertor cassette (based on the ITER 1998 design) was successfully proved during test campaigns carried out in the Divertor Refurbishment Platform (DRP) at the ENEA research centre at Brasimone over the period 1999-2003. Due to subsequent major modifications in the ITER divertor cassette design, the main focus over the past few years has been on the design and manufacture of the various components, devices and tools needed for refurbishment of the new ITER 2001 Divertor Cassette. The design of this new cassette differs substantially from the earlier version: in particular the shape, weight and attachment system of the Plasma Facing Components (PFC's) has been completely revised, and this also entailed a review of the procedures adopted for its refurbishment. One of the major requirements of the cassette refurbishment process is removal and replacement of the three PFC's. In the old cassette concept, target replacement was performed by means of a purpose-built '' C '' frame slung from a standard bridge crane. The 2001 cassette design precludes such handling methods for a number of reasons, notably because of the extremely tight inter-PFC clearances, and the need for controlled inclination of the target in addition to normal translational movements, both impossible with a simple Cartesian crane. To demonstrate the refurbishment feasibility operations for the new ITER Divertor 2001 cassettes, an experimental machine known as the Plasma-Facing Component Transporter (PFCT) has been designed, fabricated and commissioned in the years 2004-5. This full six degree-of-freedom system has been designed to handle payloads of up to 5 tonnes with good positional accuracy, and axes capable of very low joint velocities, including inclination of the PFC's over the range of ± 10 o in both horizontal axes, and controlled rotation about the vertical axis. Preliminary trials carried out during the commissioning phase have proved its

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  8. Main Quality Attributes of Monoclonal Antibodies and Effect of Cell Culture Components

    Science.gov (United States)

    Torkashvand, Fatemeh; Vaziri, Behrouz

    2017-05-01

    The culture media optimization is an inevitable part of upstream process development in therapeutic monoclonal antibodies (mAbs) production. The quality by design (QbD) approach defines the assured quality of the final product through the development stage. An important step in QbD is determination of the main quality attributes. During the media optimization, some of the main quality attributes such as glycosylation pattern, charge variants, aggregates, and low-molecular-weight species, could be significantly altered. Here, we provide an overview of how cell culture medium components affects the main quality attributes of the mAbs. Knowing the relationship between the culture media components and the main quality attributes could be successfully utilized for a rational optimization of mammalian cell culture media for industrial mAbs production.

  9. The Logical-Semantic Basis for Formation of Main Components of Enterprise Strategy

    Directory of Open Access Journals (Sweden)

    Polyakova Yana O.

    2016-11-01

    Full Text Available The modern interpretation of the essence of enterprise strategy implies transformation of internal characteristics of the enterprise into key success factors in accordance with conditions of the functioning of external environment in order to ensure its sustainable leadership position in the long term and is one of the main elements of the system of enterprise strategic management in today’s business environment. The result of the conducted research is the improvement of the logical-semantic structure of main components of enterprise strategy through the implementation of research by the inductive method, which allowed to determine the logical sequence of stages of forming key success factors, reveal the conditions for transformation of main elements of each stage and conduct a semantic evaluation of the structure of main components of enterprise strategy at each individual stage to implement the processes of goal setting and evaluate results of the implementation of the chosen business strategy.

  10. Design, construction, qualification and reliability of main components, from the safety aspect

    International Nuclear Information System (INIS)

    Crette, J.P.

    1982-01-01

    In FRANCE, the design and construction of reliable components, which condition the safe operation and availability of breeder plants, is based on the experience acquired during the operation of RAPSODIE, PHENIX and the various test facilities. The technical progress achieved on all main components is illustrated by examples taken from the CREYS-MALVILLE plant. In parallel with the development of these components, an extensive program covering research, development and the definition of design, construction and inspection rules, together with scheduling and quality assurance methods, prepares the industrialization of this reactor system, in compliance with the rules and recommendations issued by the pertinent safety authorities

  11. Analysis of Iterated Hard Decision Decoding of Product Codes with Reed-Solomon Component Codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2007-01-01

    Products of Reed-Solomon codes are important in applications because they offer a combination of large blocks, low decoding complexity, and good performance. A recent result on random graphs can be used to show that with high probability a large number of errors can be corrected by iterating...... minimum distance decoding. We present an analysis related to density evolution which gives the exact asymptotic value of the decoding threshold and also provides a closed form approximation to the distribution of errors in each step of the decoding of finite length codes....

  12. Modern technical diagnostic system for the main components of powerful turbine generator

    International Nuclear Information System (INIS)

    Ezovit, G.P.; Uglyarenko, V.P.; Burlaka, S.I.; Goroz, N.I.; Orinin, S.E.; Komaritsa, V.N.; Zav'yalov, D.N.; Mazurenko, O.A.

    2011-01-01

    The modern diagnostic system to monitor the technical state of a powerful turbine generator is considered. This system permits the detection of defects in its main components and cooling system at the early stage of their development, prevention of damage and, as a consequence, emergency shutdown of nuclear power units

  13. Remote maintenance development for ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Shibanuma, Kiyoshi

    1998-01-01

    This paper describes the overall ITER remote maintenance design concept developed mainly for in-vessel components such as diverters and blankets, and outlines the ITER R and D program to develop remote handling equipment and radiation hard components. Reactor structures inside the ITER cryostat must be maintained remotely due to DT operation, making remote handling technology basic to reactor design. The overall maintenance scenario and design concepts have been developed, and maintenance design feasibility, including fabrication and testing of full-scale in-vessel remote maintenance handling equipment and tool, is being verified. (author)

  14. Remote maintenance development for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Shibanuma, Kiyoshi

    1998-04-01

    This paper describes the overall ITER remote maintenance design concept developed mainly for in-vessel components such as diverters and blankets, and outlines the ITER R and D program to develop remote handling equipment and radiation hard components. Reactor structures inside the ITER cryostat must be maintained remotely due to DT operation, making remote handling technology basic to reactor design. The overall maintenance scenario and design concepts have been developed, and maintenance design feasibility, including fabrication and testing of full-scale in-vessel remote maintenance handling equipment and tool, is being verified. (author)

  15. Service Life Of Main Piping Component Due To Low Thermal Stresses.Fatigue

    International Nuclear Information System (INIS)

    Miroshnik, R.; Jeager, A.; Ben Haim, H.

    1998-01-01

    The paper deals with estimating the service life of the power station Main piping component and describing the repair process for extending of its service life. After a long period of service, several circular fatigue cracks have been discovered at the bottom of the Main piping component chamber. Finite element analyses of transient thermal stresses, caused by power station startup, are carried out in the paper. The calculation results show good agreement between the theoretical locations of the maximum stresses and the actual locations of the cracks. There is a good agreement between theoretical evaluation and actual service life, as well. The possibility of machining out the cracks in order to prevent their growing is examined here. The machining enables us to extend the power station component's life service

  16. [Analysis of the main components of inner ear antigens inducing autoimmune Meniere's disease in guinea pigs].

    Science.gov (United States)

    Lu, Ling; Tan, Chang-Qiang; Cui, Yu-Gui; Ding, Gui-Peng; Ju, Xiao-Bin; Li, Yu-Jin; Cai, Wen-Jun

    2008-08-01

    To investigate the main components of inner ear antigens inducing autoimmune Meniere's disease (AIMD) in guinea pigs. The guinea pigs were immunized with isologous crude inner ear antigens (ICIEAg). Then, the hearing function was measured with auditory brainstem response (ABR), the vestibular function was measured with electronystagmography (including spontaneous nystagmus and caloric test), and inner ear histopathological changes were observed by inner ear celloidin section with haematoxylin-eosin staining and observed under light microscope. According to these results, the AIMD-model animals from non-AIMD-model ones were distinguished. The special antibodies against ICIEAg in sera were measured with ELISA. The antigen-antibody reactions against different components of ICIEAg were detected by Western blotting with sera of AIMD and non-AIMD guinea pigs respectively. Then, we analysed the contrast between them and found the main components of the ICIEAg that were positive reaction in AIMD guinea pigs and negative reaction in non-AIMD guinea pigs. The result of ELISA demonstrated that the sera of both the AIMD and non-AIMD guniea pigs contained the special antibodies against ICIEAg after immunized with ICIEAg. The difference of the amount of antibody against ICIEAg between AIMD guinea pig group and non-AIMD guinea pig group was not significant. Western blotting assay showed only the sera of AIMD guinea pig contained the antibodies against the specific antigens with the molecular of 68 000, 58 000, 42 000 and 28 000. ICIEAg contain many different components, the AIMD might only happen in the guinea pigs in which the special immunization against the main components that could induce this kind of disorder appeared. The inner ear antigens with molecular of 68 000, 58 000, 42 000 and 28 000 might be the main components inducing AIMD in guinea pigs.

  17. R and D of the key components for ITER magnet supports

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

  18. R and D of the key components for ITER magnet supports

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  19. Service life monitoring of the main components at the Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Hahn, J.; Vincour, D.

    2007-01-01

    Knowledge and experience gained from the introduction and periodical implementation of life assessment of the major components of the Temelin nuclear power plant is summarized. The initial Soviet technical design of the plant did not incorporate lifetime monitoring and evaluation, therefore it was completed with demonstrative strength and lifetime calculations from Czech companies. Moreover, a Westinghouse primary circuit diagnosis and monitoring system, including the monitoring of temperature and pressure cycles for low-cycle fatigue evaluation, was installed at the plant. The DIALIFE code for the calculation of mainly the low-cycle fatigue of the key pressure components, was developed and installed subsequently as a superstructure to the monitoring system. (author)

  20. Experience with the source evaluation board method of procuring technical components for the Fermilab Main Injector

    International Nuclear Information System (INIS)

    Harding, D.J.; Collins, J.P.; Kobliska, G.R.; Chester, N.S.; Pewitt, E.G.; Fowler, W.B.

    1993-01-01

    Fermilab has adopted the Source Evaluation Board (SEB) method for procuring certain major technical components of the Fermilab Main Injector. The SEB procedure is designed to ensure the efficient and effective expenditure of Government funds at the same time that it optimizes the opportunity for attainment of project objectives. A qualitative trade-off is allowed between price and technical factors. The process involves a large amount of work and is only justified for a very limited number of procurements. Fermilab has gained experience with the SEB process in awarding subcontracts for major subassemblies of the Fermilab Main Injector dipoles

  1. Non-destructive local determination of doping additions and main components in single crystals

    International Nuclear Information System (INIS)

    Ehksperiandova, L.P.; Blank, A.B.; Kukhtina, N.N.; Afanasiadi, L.I.

    1994-01-01

    Procedures for local non-destructive determination of elements in optical and scintillation single crystals are developed. They are applied for determination of the main components (in cadmium tungstate) and doping additions (tellurium in zinc selenide, europium in gadolinium silicate). The metrological characteristics of the developed micro-analysis methods are estimated. Segregation of the main components and doping additions in the objects under consideration are investigated. Tellurium is found to be distributed uniformly on the cross-sections of bulk zinc selenide single crystals. The segregation of europium along gadolinium silicate ingots is almost absent. On the cross-section surface of cadmium tungstate single crystals the microregions are found characterized by the prevailing contents of cadmium or tungsten

  2. A conceptual design of main components sizing for UMT PHEV powertrain

    Science.gov (United States)

    Haezah, M. N.; Norbakyah, J. S.; Atiq, W. H.; Salisa, A. R.

    2015-12-01

    This paper presents a conceptual design of main components sizing for Universiti Malaysia Terengganu plug-in hybrid electric vehicle (UMT PHEV) powertrain. In the design of hybrid vehicles, it is important to identify a proper component sizes. Component sizing significantly affects vehicle performance, fuel economy and emissions. The proposed UMT PHEV has only one electric machine (EM) which functions as either a motor or generator at a time and using batteries and ultracapacitors as an energy storage system (ESS). In this work, firstly, energy and power requirements based on parameters, specifications and performance requirements of vehicle are calculated. Then, the parameters for internal combustion engine, EM and ESS are selected based on the developed Kuala Terengganu drive cycle. The results obtained from this analysis are within reasonable range and satisfactory.

  3. Correlation between magnetic field quality and mechanical components of the Large Hadron Collider main dipoles

    International Nuclear Information System (INIS)

    Bellesia, B.

    2006-12-01

    The 1234 superconducting dipoles of the Large Hadron Collider, working at a cryogenic temperature of 1.9 K, must guarantee a high quality magnetic field to steer the particles inside the beam pipe. Magnetic field measurements are a powerful way to detect assembly faults that could limit magnet performances. The aim of the thesis is the analysis of these measurements performed at room temperature during the production of the dipoles. In a large scale production the ideal situation is that all the magnets produced were identical. However all the components constituting a magnet are produced with certain tolerance and the assembly procedures are optimized during the production; due to these the reality drifts away from the ideal situation. We recollected geometrical data of the main components (superconducting cables, coil copper wedges and austenitic steel coil collars) and coupling them with adequate electro-magnetic models we reconstructed a multipolar field representation of the LHC dipoles defining their critical components and assembling procedures. This thesis is composed of 3 main parts: 1) influence of the geometry and of the assembling procedures of the dipoles on the quality of the magnetic field, 2) the use of measurement performed on the dipoles in the assembling step in order to solve production issues and to understand the behaviour of coils during the assembling step, and 3) a theoretical study of the uncertain harmonic components of the magnetic field in order to assess the dipole production

  4. Correlation between magnetic field quality and mechanical components of the Large Hadron Collider main dipoles

    Energy Technology Data Exchange (ETDEWEB)

    Bellesia, B

    2006-12-15

    The 1234 superconducting dipoles of the Large Hadron Collider, working at a cryogenic temperature of 1.9 K, must guarantee a high quality magnetic field to steer the particles inside the beam pipe. Magnetic field measurements are a powerful way to detect assembly faults that could limit magnet performances. The aim of the thesis is the analysis of these measurements performed at room temperature during the production of the dipoles. In a large scale production the ideal situation is that all the magnets produced were identical. However all the components constituting a magnet are produced with certain tolerance and the assembly procedures are optimized during the production; due to these the reality drifts away from the ideal situation. We recollected geometrical data of the main components (superconducting cables, coil copper wedges and austenitic steel coil collars) and coupling them with adequate electro-magnetic models we reconstructed a multipolar field representation of the LHC dipoles defining their critical components and assembling procedures. This thesis is composed of 3 main parts: 1) influence of the geometry and of the assembling procedures of the dipoles on the quality of the magnetic field, 2) the use of measurement performed on the dipoles in the assembling step in order to solve production issues and to understand the behaviour of coils during the assembling step, and 3) a theoretical study of the uncertain harmonic components of the magnetic field in order to assess the dipole production.

  5. Experimental activity on the definition of acceptance criteria for the ITER divertor plasma facing components

    International Nuclear Information System (INIS)

    Escourbiac, F.; Constans, S.; Vignal, N.; Cantone, V.; Richou, M.; Durocher, A.; Riccardi, B.; Bobin, I.; Jouvelot, J.L.; Merola, M.

    2009-01-01

    Tens of thousands of armor/heat sink joints will be produced by the industry during the manufacturing of ITER divertor PFC, statistically, there is a probability that joints with defects be delivered. The purpose of this paper is to study the detection and evolution during operation of calibrated defects artificially implemented on samples, as an experimental basis for the definition of acceptance criteria for the bond armor/heat sink in the frame of industrial manufacturing conditions.It was found that current CFC monoblock design option was compatible with the heat loads specified at the lower part of the vertical target (up to 20 MW/m 2 ), including the presence of armor/heat sink defects (up to 50 deg. extension for a location at 0 deg. or 45 deg.) detectable with NDE techniques developed in Europe (US, SATIR). The current W monoblock design appeared suitable for the upper part of the vertical target with defects extension up to 50 deg. but is not adapted for heat flux of 20 MW/m 2 . The studied W flat tile design proved to be compatible with fluxes of 5 MW/m 2 but unable to sustain cycling fluxes of 10 MW/m 2 .

  6. The feasibility of beryllium as structural material for the ITER plasma-facing components (PFC)

    International Nuclear Information System (INIS)

    Vieider, G.; Cardella, A.; Gorenflo, H.

    1993-01-01

    Be as plasma-facing armour has attractive features including excellent plasma compatibility, no T-retention via co-deposition and the potential for in-situ repair via plasma spraying. In order to avoid the bonding of the Be-armour to a heatsink structure in e.g., Cu-alloys, the ITER Joint Central Team (JCT) proposed for the divertor tubular elements with monolithic Be, both as plasma-facing and structural material. The analysis of these Be-tubes with 5 mm wall thickness at a heat load of 5 MW/m 2 showed that even for the most favourable assumptions thermal stresses exceed by far the allowed values according to design codes. Damage by neutrons and disruptions would worsen further the case for Be as monolithic plasma-facing and structural material. For PFC at heat flux significantly above 1 MW/m 2 it appears evident that Be should be used merely as armour bonded to a suitable structural material as heatsink. (orig.)

  7. A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F.; Missirlian, M.; Schlosser, J. [Association EURATOM-CEA Cadarache, Departement de Recherches sur la Fusion Controlee, 13 - Saint Paul lez Durance (France); Bobin-Vastra, I. [AREVA Centre Technique de Framatome, 71 - Le Creusot (France); Kuznetsov, V. [Efremov Institute, Doroga na Metallostroy, St. Petersburg (Russian Federation); Schedler, B. [Plansee AG, Reutte (Austria)

    2004-07-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m{sup 2} with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success on 2 identical mock-ups, based on this concept, that were tested in 2 different electron gun facilities. Each mock-up consisted of a CuCrZr heat sink armored with 25 flat tiles of the 3D carbon fibre composite material SEPcarb NS31 assembled with pure copper by active metal casting (AMC). The AMC tiles were electron beam welded on the CuCrZr bar, fins and slots on the neutral beam JET design were machined into the bar, then the bar was closed with a thick CuCrZr rear plug including hydraulic connections then the bar was electron beam welded onto the sidewalls. The testing results show that full ITER design specifications were achieved with margins, the critical heat flux limit was even higher than 30 MW/m{sup 2}. These results highlight the high potential of this technology for ITER divertor application.

  8. A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology

    International Nuclear Information System (INIS)

    Escourbiac, F.; Missirlian, M.; Schlosser, J.; Bobin-Vastra, I.; Kuznetsov, V.; Schedler, B.

    2004-01-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m 2 with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success on 2 identical mock-ups, based on this concept, that were tested in 2 different electron gun facilities. Each mock-up consisted of a CuCrZr heat sink armored with 25 flat tiles of the 3D carbon fibre composite material SEPcarb NS31 assembled with pure copper by active metal casting (AMC). The AMC tiles were electron beam welded on the CuCrZr bar, fins and slots on the neutral beam JET design were machined into the bar, then the bar was closed with a thick CuCrZr rear plug including hydraulic connections then the bar was electron beam welded onto the sidewalls. The testing results show that full ITER design specifications were achieved with margins, the critical heat flux limit was even higher than 30 MW/m 2 . These results highlight the high potential of this technology for ITER divertor application

  9. Identification of receptors of main sex-pheromone components of three Lepidopteran species.

    Science.gov (United States)

    Mitsuno, Hidefumi; Sakurai, Takeshi; Murai, Masatoshi; Yasuda, Tetsuya; Kugimiya, Soichi; Ozawa, Rika; Toyohara, Haruhiko; Takabayashi, Junji; Miyoshi, Hideto; Nishioka, Takaaki

    2008-09-01

    Male moths discriminate conspecific female-emitted sex pheromones. Although the chemical components of sex pheromones have been identified in more than 500 moth species, only three components in Bombyx mori and Heliothis virescens have had their receptors identified. Here we report the identification of receptors for the main sex-pheromone components in three moth species, Plutella xylostella, Mythimna separata and Diaphania indica. We cloned putative sex-pheromone receptor genes PxOR1, MsOR1 and DiOR1 from P. xylostella, M. separata and D. indica, respectively. Each of the three genes was exclusively expressed with an Or83b orthologous gene in male olfactory receptor neurons (ORNs) that are surrounded by supporting cells expressing pheromone-binding-protein (PBP) genes. By two-electrode voltage-clamp recording, we tested the ligand specificity of Xenopus oocytes co-expressing PxOR1, MsOR1 or DiOR1 with an OR83b family protein. Among the seven sex-pheromone components of the three moth species, the oocytes dose-dependently responded only to the main sex-pheromone component of the corresponding moth species. In our study, PBPs were not essential for ligand specificity of the receptors. On the phylogenetic tree of insect olfactory receptors, the six sex-pheromone receptors identified in the present and previous studies are grouped in the same subfamily but have no relation with the taxonomy of moths. It is most likely that sex-pheromone receptors have randomly evolved from ancestral sex-pheromone receptors before the speciation of moths and that their ligand specificity was modified by mutations of local amino acid sequences after speciation.

  10. Feasibility study of the cut and weld operations by RH on the cooling pipes of ITER NB components

    International Nuclear Information System (INIS)

    Pineiro, Oscar; Fernandez, Carlos; Medrano, Mercedes; Liniers, Macarena; Botija, Jose; Alonso, Javier; Sarasola, Xabier; Damiani, Carlo

    2009-01-01

    The maintenance operations of ITER NB components inside the vessel - Beam Line Components (BLC's) involve the removal of the faulty component, its transport to the hot cell as well as the reverse operations of transport of the repaired/new component and its reinstallation inside the vessel. Prior to the removal of the BLC's the cooling pipes must be detached from the component following a procedure that applies to the cutting of the pipes and subsequent welding when the component is re-installed. The purpose of this study, conducted in the framework of EFDA, is to demonstrate the feasibility of the cut and weld operations on the water pipes of the BLC's using fully remote handling techniques. Viable technologies for the cut and weld operations have been identified within the study; in particular the following aspects will be presented in the paper: - Different strategies can be pursued in the detachment of the components depending on the number of cut and weld operations to be performed on the pipes. The selected strategy will impact on the procedure to be followed likewise on important aspects as the requirements of the flexible joints assembled on the pipes. - The existing cutting techniques have been examined in the light of the remotely performed pipe cutting at the NB cell. Modifications of commercial tools have been proposed in order to adapt them to the BLC's pipes requirements. The debris produced during the cutting process must be controlled and collected, therefore a cleaning system has been integrated in the adapted cutting tool referred above. - The existing welding techniques have been also examined and compared based on different criteria such as complexity, reliability, alignment tolerances, etc. TIG welding is the preferred technique as it stands out for its superior performance. The commercial tools identified need to be adapted to the NB environment. - The alignment of the pipes is a critical issue concerning the remote welding. A proper alignment

  11. RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Pt. II. Analysis of ITER plasma facing components

    International Nuclear Information System (INIS)

    Federici, G.; Raffray, A.R.

    1997-01-01

    For pt.I see ibid., p.85-100, 1997. The transient thermal model RACLETTE (acronym of Rate Analysis Code for pLasma Energy Transfer Transient Evaluation) described in part I of this paper is applied here to analyse the heat transfer and erosion effects of various slow (100 ms-10 s) high power energy transients on the actively cooled plasma facing components (PFCs) of the International Thermonuclear Experimental Reactor (ITER). These have a strong bearing on the PFC design and need careful analysis. The relevant parameters affecting the heat transfer during the plasma excursions are established. The temperature variation with time and space is evaluated together with the extent of vaporisation and melting (the latter only for metals) for the different candidate armour materials considered for the design (i.e., Be for the primary first wall, Be and CFCs for the limiter, Be, W, and CFCs for the divertor plates) and including for certain cases low-density vapour shielding effects. The critical heat flux, the change of the coolant parameters and the possible severe degradation of the coolant heat removal capability that could result under certain conditions during these transients, for example for the limiter, are also evaluated. Based on the results, the design implications on the heat removal performance and erosion damage of the various ITER PFCs are critically discussed and some recommendations are made for the selection of the most adequate protection materials and optimum armour thickness. (orig.)

  12. RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part II: Analysis of ITER plasma facing components

    Science.gov (United States)

    Federici, Gianfranco; Raffray, A. René

    1997-04-01

    The transient thermal model RACLETTE (acronym of Rate Analysis Code for pLasma Energy Transfer Transient Evaluation) described in part I of this paper is applied here to analyse the heat transfer and erosion effects of various slow (100 ms-10 s) high power energy transients on the actively cooled plasma facing components (PFCs) of the International Thermonuclear Experimental Reactor (ITER). These have a strong bearing on the PFC design and need careful analysis. The relevant parameters affecting the heat transfer during the plasma excursions are established. The temperature variation with time and space is evaluated together with the extent of vaporisation and melting (the latter only for metals) for the different candidate armour materials considered for the design (i.e., Be for the primary first wall, Be and CFCs for the limiter, Be, W, and CFCs for the divertor plates) and including for certain cases low-density vapour shielding effects. The critical heat flux, the change of the coolant parameters and the possible severe degradation of the coolant heat removal capability that could result under certain conditions during these transients, for example for the limiter, are also evaluated. Based on the results, the design implications on the heat removal performance and erosion damage of the variuos ITER PFCs are critically discussed and some recommendations are made for the selection of the most adequate protection materials and optimum armour thickness.

  13. Methodology on ageing management review for main components of a PWR NPP

    International Nuclear Information System (INIS)

    Dou Yikang; He Yinbiao; Xu Xuelian; Zhang Ming; Liang Xingyun

    2005-01-01

    According to the requirements of NNSA, for Chinese operational NPPs periodical safety review (PSR) should be carried out every 10 years. Ageing management is one of the important safety factors to be reviewed. Entrusted by Qinshan Nuclear Power Plant (QNPP), Shanghai Nuclear Engineering Research and Design Institute (SNERDI) carried out the ageing management review (AMR), as a part of the first PSR of QNPP, from 2001 to 2003. This paper summarizes the methodology of the AMR process, including screening of critical components and structures, identification of main ageing mechanisms and their indicators and the tabulated review process, etc. 15 components and structures, hereafter referred as equipment, were selected as review objects based on their significance of safety, replaceability and cost-benefit considerations. To these objects, the main ageing mechanisms and relevant ageing indicators were identified according to specific working and environmental condition, design and manufacture information, operation and maintenance history, etc. The review can be divided into two parallel parts, the review for specific equipment and the review for overall management procedures and their implementation. To typical components, such as RPV and SG, fatigue analysis based on operational transient accounting was carried out to observe the actual safety margins. Through the review, the weaknesses in ageing management and potential threats to structural integrity were identified and thus continued improvement can be made in the next period of 10 years. (authors)

  14. Identification and Hierarchy of Main Electronic Health Record Components in Occupational Medicine

    Directory of Open Access Journals (Sweden)

    Dorin TRIFF

    2010-12-01

    Full Text Available Starting from the legal requirements relating to structuring of medical records in occupational medicine and international requirements regarding the certification of electronic health records we have focused on structuring and then evaluating an EHR model in occupational medicine that integrates the main functions and certification criteria required by the European and US certification bodies. The application we designed, called Medmun, structured for use in occupational medicine practices based on the model of medical file provided by the Romanian legislation, integrates both necessary components of occupational medicine practice for administration of characteristic information related to socio-economic unit, work place, health surveillance as well as components of specific EHR functionality. The application has been submitted for free evaluation by specialist physicians of five counties over a period of nine months and subsequently assessed using a questionnaire on the usefulness of specific functional components in the EHR occupational medicine practice. The model was positively evaluated after experimental employment by occupational health practitioners. They consider that absence of legislative support for EHR implementation in medical practice is the main obstacle to the use of such applications in occupational medicine practice.

  15. [Simultaneous determination of five main index components and specific chromatograms analysis in Xiaochaihu granules].

    Science.gov (United States)

    Zhuang, Yan-Shuang; Cai, Hao; Liu, Xiao; Cai, Bao-Chang

    2012-01-01

    Reversed phase high performance liquid chromatography with diode array detector was employed for simultaneous determination of five main index components and specific chromatograms analysis in Xiaochaihu granules with a linear gradient elution of acetonitrile-water (containing 0.1% phosphoric acid) as mobile phase. The results showed that five main index components (baicalin, baicalein, wogonoside, wogonin, enoxolone) were separated well under the analytical condition. The linear ranges of five components were 0.518 - 16.576, 0.069 - 2.197, 0.167 - 5.333, 0.009 - 0.297 and 0.006 - 0.270 mg x g(-1), respectively. The correlation coefficients were 0.999 9, and the average recoveries ranged from 95% to 105%. Twelve common peaks were selected as the specific chromatograms of Xiaochaihu granules with baicalin as the reference peak. There were good similarities between the reference and the ten batches of samples. The similarity coefficients were no less than 0.9. The analytical method established is highly sensitive with strong specificity and it can be used efficiently in the quality control of Xiaochaihu granules.

  16. Simultaneous quantitative analysis of main components in linderae reflexae radix with one single marker.

    Science.gov (United States)

    Wang, Li-Li; Zhang, Yun-Bin; Sun, Xiao-Ya; Chen, Sui-Qing

    2016-05-08

    Establish a quantitative analysis of multi-components by the single marker (QAMS) method for quality evaluation and validate its feasibilities by the simultaneous quantitative assay of four main components in Linderae Reflexae Radix. Four main components of pinostrobin, pinosylvin, pinocembrin, and 3,5-dihydroxy-2-(1- p -mentheneyl)- trans -stilbene were selected as analytes to evaluate the quality by RP-HPLC coupled with a UV-detector. The method was evaluated by a comparison of the quantitative results between the external standard method and QAMS with a different HPLC system. The results showed that no significant differences were found in the quantitative results of the four contents of Linderae Reflexae Radix determined by the external standard method and QAMS (RSD <3%). The contents of four analytes (pinosylvin, pinocembrin, pinostrobin, and Reflexanbene I) in Linderae Reflexae Radix were determined by the single marker of pinosylvin. This fingerprint was the spectra determined by Shimadzu LC-20AT and Waters e2695 HPLC that were equipped with three different columns.

  17. HELCZA-High heat flux test facility for testing ITER EU first wall components.

    Czech Academy of Sciences Publication Activity Database

    Prokůpek, J.; Samec, K.; Jílek, R.; Gavila, P.; Neufuss, S.; Entler, Slavomír

    2017-01-01

    Roč. 124, November (2017), s. 187-190 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : HELCZA * High heat flux * Electron beam testing * Test facility * Plasma facing components * First wall * Divertora Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 www.sciencedirect.com/science/article/pii/S0920379617302818

  18. Dynamic models of reduced order of main components of a M SR

    International Nuclear Information System (INIS)

    Garcia B, F. B.; Morales S, J. B.; Polo L, M. A.; Espinosa P, G.

    2011-11-01

    The reactors of melted salts called Molten Salt Fast Reactor (MSFR), have seen a resurgence of interest in the last decade. This design is one of the six proposed for the IV generation reactors. The most active development was in the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratories (ORNL). In this work the mathematician modeling of the main components in the primary and secondary circuits of a M SR is presented. In particular the dynamics of the heat exchanger is analyzed and they are considered several materials to optimize the system thermodynamically. (Author)

  19. Radiation effects on optical components of a laser radar sensor designed for remote metrology in ITER

    International Nuclear Information System (INIS)

    Menon, M.M.; Grann, E.B.; Slotwinski, A.

    1997-09-01

    A frequency modulated laser radar is being developed for in-vessel metrology and viewing of plasma-facing surfaces. Some optical components of this sensor must withstand intense gamma radiation (3 x 10 6 rad/h) during operation. The authors have tested the effect of radiation on a silica core polarization maintaining optical fiber and on TeO 2 crystals at doses up to ∼ 10 9 rad. Additional tests are planned for evaluating the performance of a complete acousto-optic (AO) scanning device. The progress made in these tests is also described

  20. Selenium enrichment on Cordyceps militaris link and analysis on its main active components.

    Science.gov (United States)

    Dong, Jing Z; Lei, C; Ai, Xun R; Wang, Y

    2012-03-01

    To investigate the effects of selenium on the main active components of Cordyceps militaris fruit bodies, selenium-enriched cultivation of C. militaris and the main active components of the fruit bodies were studied. Superoxide dismutase (SOD) activity and contents of cordycepin, cordycepic acid, and organic selenium of fruit bodies were sodium selenite concentration dependent; contents of adenosine and cordycep polysaccharides were significantly enhanced by adding sodium selenite in the substrates, but not proportional to sodium selenite concentrations. In the cultivation of wheat substrate added with 18.0 ppm sodium selenite, SOD activity and contents of cordycepin, cordycepic acid, adenosine, cordycep polysaccharides, and total amino acids were enhanced by 121/145%, 124/74%, 325/520%, 130/284%, 121/145%, and 157/554%, respectively, compared to NS (non-selenium-cultivated) fruit bodies and wild Cordyceps sinensis; organic selenium contents of fruit bodies reached 6.49 mg/100 g. So selenium-enriched cultivation may be a potential way to produce more valuable medicinal food as a substitute for wild C. sinensis.

  1. 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.)

  2. ITER overview

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  3. ITER Overview

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  4. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10 6 R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  5. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan)] [and others

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10{sup 6} R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  6. Low cycle thermal fatigue testing of beryllium grades for ITER plasma facing components

    International Nuclear Information System (INIS)

    Watson, R.D.; Youchison, D.L.; Dombrowski, D.E.; Guiniatouline, R.N.; Kupriynov, I.B.

    1996-01-01

    A novel technique has been used to test the relative low cycle thermal fatigue resistance of different grades of US and Russian beryllium, which is proposed as plasma facing armor for fusion reactor first wall, limiter, and divertor components. The 30 kW electron beam test system at Sandia National Laboratories was used to sweep the beam spot along one direction at 1 Hz. This produces a localized temperature ''spike'' of 750 degree C for each pass of the beam. Large thermal stresses in excess of the yield strength are generated due to very high spot heat flux, 250 MW/m 2 . Cyclic plastic strains on the order of 0.6% produced visible cracking on the heated surface in less than 3000 cycles. An in-vacuo fiber optic borescope was used to visually inspect the beryllium surfaces for crack initiation. Grades of US beryllium tested included: S-65C, S- 65H, S-200F, S-200F-H, SR-200, I-400, extruded high purity, HIP'd spherical powder, porous beryllium (94% and 98% dense), Be/30% BeO, Be/60% BeO, and TiBe 12 . Russian grades included: TGP-56, TShGT, DShG-200, and TShG-56. Both the number of cycles to crack initiation, and the depth of crack propagation, were measured. The most fatigue resistant grades were S-65C, DShG-200, TShGT, and TShG-56. Rolled sheet Be (SR-200) showed excellent crack propagation resistance in the plane of rolling, despite early formation of delamination cracks. Only one sample showed no evidence of surface melting, Extruded (T). Metallographic and chemical analyses are provided. Good agreement was found between the measured depth of cracks and a 2-D elastic-plastic finite element stress analysis

  7. Feasibility study of the cut and weld operations by RH on the cooling pipes of ITER NB components

    Energy Technology Data Exchange (ETDEWEB)

    Pineiro, Oscar; Fernandez, Carlos [TECNATOM Avda. Montes de Oca 28700 S Sebastian de los Reyes, Madrid (Spain); Medrano, Mercedes [EURATOM-CIEMAT Association for Fusion. Avda. Complutense, 22. 28040 Madrid (Spain)], E-mail: mercedes.medrano@ciemat.es; Liniers, Macarena; Botija, Jose; Alonso, Javier; Sarasola, Xabier [EURATOM-CIEMAT Association for Fusion. Avda. Complutense, 22. 28040 Madrid (Spain); Damiani, Carlo [EFDA-Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2009-06-15

    The maintenance operations of ITER NB components inside the vessel - Beam Line Components (BLC's) involve the removal of the faulty component, its transport to the hot cell as well as the reverse operations of transport of the repaired/new component and its reinstallation inside the vessel. Prior to the removal of the BLC's the cooling pipes must be detached from the component following a procedure that applies to the cutting of the pipes and subsequent welding when the component is re-installed. The purpose of this study, conducted in the framework of EFDA, is to demonstrate the feasibility of the cut and weld operations on the water pipes of the BLC's using fully remote handling techniques. Viable technologies for the cut and weld operations have been identified within the study; in particular the following aspects will be presented in the paper: - Different strategies can be pursued in the detachment of the components depending on the number of cut and weld operations to be performed on the pipes. The selected strategy will impact on the procedure to be followed likewise on important aspects as the requirements of the flexible joints assembled on the pipes. - The existing cutting techniques have been examined in the light of the remotely performed pipe cutting at the NB cell. Modifications of commercial tools have been proposed in order to adapt them to the BLC's pipes requirements. The debris produced during the cutting process must be controlled and collected, therefore a cleaning system has been integrated in the adapted cutting tool referred above. - The existing welding techniques have been also examined and compared based on different criteria such as complexity, reliability, alignment tolerances, etc. TIG welding is the preferred technique as it stands out for its superior performance. The commercial tools identified need to be adapted to the NB environment. - The alignment of the pipes is a critical issue concerning the remote welding

  8. Chemical composition of Galla chinensis extract and the effect of its main component(s) on the prevention of enamel demineralization in vitro

    NARCIS (Netherlands)

    Huang, X.L.; Liu, M.D.; Li, J.Y.; Zhou, X.D.; ten Cate, J.M.

    2012-01-01

    To determine the chemical composition of Galla chinensis extract (GCE) by several analysis techniques and to compare the efficacy of GCE and its main component(s) in inhibition of enamel demineralization, for the development of future anticaries agents, main organic composition of GCE was

  9. Effect of humates in diet of dairy cows on the raw milk main components

    Directory of Open Access Journals (Sweden)

    Miroslava Potůčková

    2017-01-01

    Full Text Available The effect of supplemental humic substances (HS on the main milk components was investigated. A total of 10 dairy cows (Czech pied cattle, crossbred Czech pied cattle ´ Ayrshire and crossbred Czech pied cattle ´ Red Holstein were tested. Animals were randomly divided into 2 groups, control (C and experimental (E. Animals fed the same feed mixture and group E was additionally supplemented with HS (200 mg.kg-1 of product Humafit prepared from the Sakhalin Leonardite. The experimental period took 3 months. Cows were milked twice a day. The milk composition (lactose, fat, crude protein, pure protein and casein of every cow was monitored on days 0, 14, 28, 42, 56, 70 and 84 of the experiment. Pure protein content was determined by Kjeldahl method, other components were analysed using an infrared analyserMilkoScan FT 120. It was found that the crude protein, pure protein and casein content in milk of group E significantly (p <0.05 increased from the 56th day of the experimental period. Differences of the protein fraction contents in group C and of the dry matter, non-fat dry matter, lactose and fat content in both groups were non-significant (p <0.05.Higher protein and especially casein content in milk could be very important for the cheesemaking as it could increase the cheese yield. Normal 0 21 false false false CS X-NONE X-NONE

  10. Deformation Analysis of the Main Components in a Single Screw Compressor

    Science.gov (United States)

    Liu, Feilong; Liao, Xueli; Feng, Quanke; Van Den Broek, Martijn; De Paepe, Michel

    2015-08-01

    The single screw compressor is used in many fields such as air compression, chemical industry and refrigeration. During operation, different gas pressures and temperatures applied on the components can cause different degrees of deformation, which leads to a difference between the thermally induced clearance and the designed clearance. However, limited research about clearance design is reported. In this paper, a temperature measurement instrument and a convective heat transfer model were described and used to establish the temperature of a single screw air compressor's casing, screw rotor and star wheel. 3-D models of these three main components were built. The gas force deformation, thermal- structure deformation and thermal-force coupling deformation were carried out by using a finite element simulation method. Results show that the clearance between the bottom of the groove and the top of star wheel is reduced by 0.066 mm, the clearance between the side of groove and the star wheel is reduced by 0.015 mm, and the clearance between the cylinder and the rotor is reduced by 0.01 mm. It is suggested that these deformations should be taken into account during the design of these clearances.

  11. Recent progress in R&D for long pulse and ultra-high voltage components for the ITER HNB

    Energy Technology Data Exchange (ETDEWEB)

    Kashiwagi, Mieko, E-mail: kashiwagi.mieko@jaea.go.jp; Umeda, Naotaka; Kojima, Atsushi; Yoshida, Masafumi; Tobari, Hiroyuki; Dairaku, Masayuki; Yamanaka, Haruhiko; Maejima, Tetsuya; Yamashita, Yasuo; Shibata, Naoki; Watanabe, Kazuhiro; Hanada, Masaya

    2015-10-15

    Highlights: • Long-pulse acceleration of MeV-class ion beams is demonstrated in JAEA. • The pulse length is extended from 0.4 s to 60 s, successfully. • R&D of 1 MV power supply is in progressed as scheduled. • Feasibility of 1 MV insulating transformer is confirmed in the mockup. - Abstract: Toward ITER, the long-pulse acceleration of MeV-class negative ion beams has been successfully demonstrated and R&Ds and the procurement of high-voltage components of 1 MV power supply system have been progressed. In the accelerator development after the production of 1 MeV beams, an extraction grid, where a water cooling is reinforced and the beam is steered by aperture offset to suppress grid heat loads, has been newly developed to extend the pulse duration of the MeV-class beams. As the result, a total grid heat load has been reduced from 20% to 10% of the total beam power. The beam pulse was successfully increased from 0.4 s to 60 s at 0.68 MeV, 100 A/m{sup 2} and 12 s at 0.89 MeV, 156 A/m{sup 2} beams. As for the procurement of the 1 MV power supply system, 1 MV insulating transformer as the critical component has been developed. In the mock-up, a dc 1.2 MV for 1 h with the margin of 20% was stably sustained. Through the R&Ds of the critical components, design of five power supplies procured by JAEA has been finalized, and out of which a testing power supply has been completed in June 2014.

  12. Influence of Cements Containing Calcareous Fly Ash as a Main Component Properties of Fresh Cement Mixtures

    Science.gov (United States)

    Gołaszewski, Jacek; Kostrzanowska-Siedlarz, Aleksandra; Ponikiewski, Tomasz; Miera, Patrycja

    2017-10-01

    The main goal of presented research was to examine usability of cements containing calcareous fly ash (W) from technological point of view. In the paper the results of tests concerning the influence of CEM II and CEM IV cements containing fly ash (W) on rheological properties, air content, setting times and plastic shrinkage of mortars are presented and discussed. Moreover, compatibility of plasticizers with cements containing fly ash (W) was also studied. Additionally, setting time and hydration heat of cements containing calcareous fly ash (W) were determined. In a broader aspect, the research contributes to promulgation of the possibility of using calcareous fly ash (W) in cement and concrete technology, what greatly benefits the environment protection (utilization of waste fly ash). Calcareous fly ash can be used successfully as the main component of cement. Cements produced by blending with processed fly ash or cements produced by interginding are characterized by acceptable technological properties. In respect to CEM I cements, cements containing calcareous fly ash worsen workability, decrease air content, delay setting time of mixtures. Cements with calcareous fly ash show good compatibility with plasticizers.

  13. Replacement of heavy components of the Main Primary System (MPS). Recent innovations made by Framatome ANP

    International Nuclear Information System (INIS)

    Chanussot, J.M.; Thevenet, R.

    2006-01-01

    The replacement of heavy components is the result of widespread stress corrosion of Inconel 600 (and alloys 82/182) in the primary system. Following the corrosion of steam generator tubes, which led to the first steam generator replacement (SGR) operations, work has begun on reactor vessel head replacements (RVHR) and pressurizer replacements, pending treatment of the dissimilar butt welds (DBW). The replacement of sections of the Main Primary System is one way of optimizing plant lifetime besides preventing stress corrosion in primary components and combating phenomena as thermal fatigue or certain metallurgical weaknesses. The increase in the number of major operations carried out in the reactor building on PWR and BWR reactor geometries has called for major technical innovations. AREVA, has over the past twenty years carried out more than forty SGRs throughout the world, including the supply of 115 steam generators of all types. This equates to over 60% of all SGR operations performed in Europe and the USA. Today, most SGR operations are performed in the USA, though some still continue in Europe and Asia. Operations have been planned until 2015. The first RVHR operations were carried out in France in 1993, and have been widespread in Europe since 1996 (Spain and Belgium), and in the USA since 2002 (Davis Besse). Since 1993 AREVA, through its subsidiaries Framatome ANP, Jeumont SA and SGT, has performed 72 RVHRs, i.e. most of those carried out in Europe and the USA. Today, RVHR operations take place throughout the world and have been planned until 2014. The first pressurizer replacement took place in the fall of 2005 in Saint Lucie, USA. Others are to follow (Fort Calhoun, Millestone). A significant number of partial or complete pressurizer replacements are now being scheduled. Special focus must be given to this new operation, as a decision is required between the following types of replacements, taking into account the number of pressurizer heaters and the

  14. Design and rescue scenario of common repair equipment for in-vessel components in ITER hot cell

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Takeda, Nobukazu; Nakahira, Masataka; Shibanuma, Kiyoshi

    2006-06-01

    Transportation of the in-vessel components to be repaired in the ITER hot cell is carried by two kinds of transporters, i.e., overhead cranes and floor vehicles. The access area for repair operations in the hot cell is duplicated by these transporters. Clear sharing of the respective roles of these transporters with the minimum duplication is therefore useful for rationalization. The overhead cranes, which are independently installed in the respective cells in the hot sell, cannot pass through the components to be repaired between cells, i.e., receiving cell and refurbishment cell as an example. If the floor vehicle with simple mechanisms can cover the inaccessible area for the overhead cranes, a global transporter system in the hot cell will be simplified and the reliability will be increased. Based on this strategy, the overhead crane and floor vehicle concepts are newly proposed. The overhead crane has an adapter for change of the end-effectors, which can be easily changed, to grasp many kinds of components to be repaired. The floor vehicle, which is equipped with wheel mechanisms for transportation, is just to pass through the components between cells with only straight (linear) motion on the floor. The simple wheel mechanism can solve the spread of the dust, which is the critical issue of the original air bearing mechanism for traveling in the 2001 FDR design. Rescue scenarios and procedures in the hot cell are also studied in this report. The proposed rescue crane has major two functions for rescue operations of the hot cell facility, i.e., one for the overhead crane and the other for refurbishment equipment such as workstation for divertor repair. The rescue of the faulty overhead crane is carried out using the rescue tool installed on the rescue crane or directly traveled by pushing/pulling by the rescue crane after docking on the faulty overhead crane. For the rescue of the workstation, the rescue crane consists of a telescopic manipulator (maximum length

  15. Analysis of heat transfer and erosion effects on ITER divertor plasma facing components induced by slow high-power transients

    International Nuclear Information System (INIS)

    Federici, G.; Raffray, A.R.; Chiocchio, S.; Esser, B.; Dietz, J.; Igitkhanov, Y.; Janeschitz, G.

    1995-01-01

    This paper presents the results of an analysis carried out to investigate the thermal response of ITER divertor plasma facing components (PFC's) clad with Be, W, and CFC, to high-recycling, high-power thermal transients (i.e. 10--30 MW/m 2 ) which are anticipated to last up to a few seconds. The armour erosion and surface melting are estimated for the different plasma facing materials (PFM's) together with the maximum heat flux to the coolant, and armour/heat-sink interface temperature. The analysis assumes that intense target evaporation will lead to high radiative power losses in the plasma in front of the target which self-protects the target. The cases analyzed clarify the influence of several key parameters such as the plasma heat flux to the target, the loss of the melt layer, the duration of the event, the thickness of the armour, and comparison is made with cases without vapor shielding. Finally, some implications for the performance and lifetime of divertor PFC's clad with different PFM's are discussed

  16. Steady state and transient thermal-hydraulic characterization of full-scale ITER divertor plasma facing components

    International Nuclear Information System (INIS)

    Tincani, A.; Malavasi, A.; Ricapito, I.; Riccardi, B.; Di Maio, P.A.; Vella, G.

    2007-01-01

    In the frame of the activities related to ITER divertor R and D, ENEA CR Brasimone was charged by EFDA (European Fusion Design Agreement) to investigate the thermal-hydraulic behaviour of the full-scale divertor plasma facing components, i.e. Inner Vertical Target, Dome Liner and Outer Vertical Target, both in steady state and during draining and drying transient. More in detail, for each PFC, the first phase of the work is the steady state hydraulic characterization which consists of: - measurements of pressure drops at different temperatures; - determination of the velocity distribution in the internal channels; - check the possible insurgence of cavitation. The subsequent phase of the thermal-hydraulic characterization foresees a testing campaign of draining and drying procedure by means of a suitable gas flow. The objective of this experimental procedure is to eliminate in the most efficient way the residual amount of water after gravity discharge. In order to accomplish this experimental campaign a significant modification of CEF1 loop has been designed and realized. This paper presents, first of all, the experimental set-up, the agreed test matrix and the achieved results for both steady state and transient tests. Moreover, the level of the implementation of a predictive hydraulic model, based on RELAP 5 code, as well as its results are described, discussed and compared with the experimental ones. (orig.)

  17. THE MAIN COMPONENTS OF THE CIVIC EDUCATION IN MODERN SCHOOL (USING THE EXAMPLE OF THE USA

    Directory of Open Access Journals (Sweden)

    Inna Stanislavovna Bessarabova

    2015-11-01

    Full Text Available The urgency of the topic is stipulated by the importance of the problem of civic education under modern conditions of forming of the constitutional state and civic society, which has the rights and freedoms of citizens as the highest value.The purpose of the research is to reveal and characterize the peculiarities of the process of civic education of schoolchildren in modern secondary school (using the example of the USA.The results of the research showed that civic education involves three basic components: civic knowledge, civic skills and civic position. Civic knowledge includes clear understanding of the government structure and the processes of law adoption. At this stage the main focus in the process of civic education is directed to the students’ knowledge and skills to analyze historical and political events in the country, to understand the directions of the state policy in the country. The main function of the civic knowledge is its stimulation of the civic activity.The essential part of civic education is the formation of the civic skills, which are dependent on many factors including home environment, active and responsible participation in political life of the society, general education. These skills are classified as interaction, control and influence.The next basic component of civic education – civic position – is connected with personal and social characteristics of a citizen, which are important for active participation under the conditions of constitutional democracy. Civic position is closely connected with conscious readiness of a person to take part in social activity, availability of the theoretical civic know-ledge and practical civic skills. Itshould be mentioned that civic knowledge, civic skills and civic position are interconnected.Practical implications of the results. The results of the research are applicable in the process of working out of the new technologies, methods and approaches in education to

  18. [Characteristic fingerprint analysis and determination of main components on Andrographis paniculata extract by UPLC].

    Science.gov (United States)

    Zhang, Hui-ye; Xu, Xiao-fei; Huang, Yong; Wamg, De-qin; Deng, Qiao-hua

    2014-06-01

    To establish an analytical method for characteristic fingerprint and determination of main components of Andrographis paniculata Extract by UPLC. The chromatographic conditions were Waters ACQUITY UPLC BEH-C18 (2. 1 mm x 0 mm,1.7 μm)by gradient elute using acetonitrile-water as mobile phase(0 -2 min,20% ~ 25% A;2 ~ 5 min,25% ~ 35% A;5 ~ 7 min,35% A;7 ~10 min,35%~ 55% A) at a flow rate of 0. 5 min/mL,detecting wavelength at 220 nm. Results:Contents of the andrographolide, neoandrographolide, 14-deoxyandrographolide and 14-deoxy-l11,12-didehydroandrographolide had good resolution with the correlation coefficients exceed 0. 9999 and the average percent recovery lied in 97. 2% to 103.9%, RSD was less than 3.0% (n = 6). The chromatograms of Andrographis paniculata Extract shared seven common peaks in which four of them were recognized by reference standard with the similarities over 0. 9. It is a fast,accurate and validated method,and can be useful in quality evaluations of Andrographis paniculata Extract.

  19. A review of Neuropharmacology Effects of Nigella sativa and Its Main Component, Thymoquinone.

    Science.gov (United States)

    Javidi, Soheila; Razavi, Bibi Marjan; Hosseinzadeh, Hossein

    2016-08-01

    Neuropharmacology is the scientific study of drug effect on nervous system. In the last few years, different natural plants and their active constituents have been used in neurological therapy. The availability, lower price, and less toxic effects of herbal medicines compared with synthetic agents make them as simple and excellent choice in the treatment of nervous diseases. Nigella sativa, which belongs to the botanical family of Ranunculaceae, is a widely used medicinal plant all over the world. In traditional and modern medicines several beneficial properties have been attributed to N. sativa and its main component, thymoquinone (TQ). In this review, various studies in scientific databases regarding the neuropharmacological aspects of N. sativa and TQ have been introduced. Results of these studies showed that N. sativa and TQ have several properties including anticonvulsant, antidepressant, anxiolytic, anti-ischemic, analgesic, antipsychotic, and memory enhancer. Furthermore, its protective effects against neurodegenerative diseases such as Alzheimer, Parkinson and multiple sclerosis have been discussed. Although there are many studies indicating the beneficial actions of this plant in nervous system, the number of research projects relating to the human reports is rare. Copyright © 2016 John Wiley & Sons, Ltd. Copyright © 2016 John Wiley & Sons, Ltd.

  20. Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

    Energy Technology Data Exchange (ETDEWEB)

    Ritz, Guillaume Henri

    2011-07-01

    In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m{sup -2} as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m{sup -2}. The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform

  1. Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

    International Nuclear Information System (INIS)

    Ritz, Guillaume Henri

    2011-01-01

    In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m -2 as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m -2 . The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform sophisticated

  2. Materials challenges for ITER - Current status and future activities

    Energy Technology Data Exchange (ETDEWEB)

    Barabash, V. [ITER International Team, Boltsmannstrasse 2, 85748 Garching (Germany)]. E-mail: valdimir.barabash@iter.org; Peacock, A. [EFDA Close Support Unit, 85748 Garching (Germany); Fabritsiev, S. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Kalinin, G. [ENES, P.O. Box 788, 101000 Moscow (Russian Federation); Zinkle, S. [Metals and Ceramics Division, ORNL, P.O. Box 2008, Oak Ridge, TN 37831-6138 (United States); Rowcliffe, A. [Metals and Ceramics Division, ORNL, P.O. Box 2008, Oak Ridge, TN 37831-6138 (United States); Rensman, J.-W. [NRG, P.O. Box 25, 1755 ZG Petten (Netherlands); Tavassoli, A.A. [Commissariat a l' Energie Atomique, CEA/Saclay, 91191 Gif sur Yvette cedex (France); Marmy, P. [CRPP, EPFL, Association EURATOM-Confederation Suisse, 5232, Villigen PSI (Switzerland); Karditsas, P.J. [EURATOM/UKAEA Fusion Association, Abingdon, Oxon OX14 3DB (United Kingdom); Gillemot, F. [AEKI Atomic Research Institute, 1121 Budapest, (Hungary); Akiba, M. [JAEA, Naka-machi, Naka-gun, Ibaraki-ken 311-0193 (Japan)

    2007-08-01

    ITER will be the first experimental fusion facility, which brings together the key physical, material and technological issues related to development of fusion reactors. The design of ITER is complete and the construction will start soon. This paper discusses the main directions of the project oriented materials activity and main challenges related to selection of materials for the ITER components. For each application in ITER the main materials issues were identified and these issues were addressed in the dedicated ITER R and D program. The justification of materials performance was fully documented, which allows traceability and reliability of design data. Several examples are given to illustrate the main achievements and recommendations from the recently updated ITER Materials Properties Handbook. The main ongoing and future materials activities are described.

  3. Probabilistic Structural Analysis Methods for select space propulsion system components (PSAM). Volume 2: Literature surveys of critical Space Shuttle main engine components

    Science.gov (United States)

    Rajagopal, K. R.

    1992-01-01

    The technical effort and computer code development is summarized. Several formulations for Probabilistic Finite Element Analysis (PFEA) are described with emphasis on the selected formulation. The strategies being implemented in the first-version computer code to perform linear, elastic PFEA is described. The results of a series of select Space Shuttle Main Engine (SSME) component surveys are presented. These results identify the critical components and provide the information necessary for probabilistic structural analysis. Volume 2 is a summary of critical SSME components.

  4. A network analysis of the Chinese medicine Lianhua-Qingwen formula to identify its main effective components.

    Science.gov (United States)

    Wang, Chun-Hua; Zhong, Yi; Zhang, Yan; Liu, Jin-Ping; Wang, Yue-Fei; Jia, Wei-Na; Wang, Guo-Cai; Li, Zheng; Zhu, Yan; Gao, Xiu-Mei

    2016-02-01

    Chinese medicine is known to treat complex diseases with multiple components and multiple targets. However, the main effective components and their related key targets and functions remain to be identified. Herein, a network analysis method was developed to identify the main effective components and key targets of a Chinese medicine, Lianhua-Qingwen Formula (LQF). The LQF is commonly used for the prevention and treatment of viral influenza in China. It is composed of 11 herbs, gypsum and menthol with 61 compounds being identified in our previous work. In this paper, these 61 candidate compounds were used to find their related targets and construct the predicted-target (PT) network. An influenza-related protein-protein interaction (PPI) network was constructed and integrated with the PT network. Then the compound-effective target (CET) network and compound-ineffective target network (CIT) were extracted, respectively. A novel approach was developed to identify effective components by comparing CET and CIT networks. As a result, 15 main effective components were identified along with 61 corresponding targets. 7 of these main effective components were further experimentally validated to have antivirus efficacy in vitro. The main effective component-target (MECT) network was further constructed with main effective components and their key targets. Gene Ontology (GO) analysis of the MECT network predicted key functions such as NO production being modulated by the LQF. Interestingly, five effective components were experimentally tested and exhibited inhibitory effects on NO production in the LPS induced RAW 264.7 cell. In summary, we have developed a novel approach to identify the main effective components in a Chinese medicine LQF and experimentally validated some of the predictions.

  5. Main components and characteristics of landslide early warning systems operational worldwide

    Science.gov (United States)

    Piciullo, Luca; Cepeda, José

    2017-04-01

    During the last decades the number of victims and economic losses due to natural hazards are dramatically increased worldwide. The reason can be mainly ascribed to climate changes and urbanization in areas exposed at high level of risk. Among the many mitigation measures available for reducing the risk to life related to natural hazards, early warning systems certainly constitute a significant cost-effective option available to the authorities in charge of risk management and governance. The aim is to help and protect populations exposed to natural hazards, reducing fatalities when major events occur. Landslide is one of the natural hazards addressed by early warning systems. Landslide early warning systems (LEWSs) are mainly composed by the following four components: set-up, correlation laws, decisional algorithm and warning management. Within this framework, the set-up includes all the preliminary actions and choices necessary for designing a LEWS, such as: the area covered by the system, the types of landslides and the monitoring instruments. The monitoring phase provides a series of important information on different variables, considered as triggering factors for landslides, in order to define correlation laws and thresholds. Then, a decisional algorithm is necessary for defining the: number of warning levels to be employed in the system, decision making procedures, and everything else system managers may need for issuing warnings in different warning zones. Finally the warning management is composed by: monitoring and warning strategy; communication strategy; emergency plan and, everything connected to the social sphere. Among LEWSs operational worldwide, two categories can be defined as a function of the scale of analysis: "local" and "territorial" systems. The scale of analysis influences several actions and aspects connected to the design and employment of the system, such as: the actors involved, the monitoring systems, type of landslide phenomena

  6. Polysaccharide components from the scape of Musa paradisiaca: main structural features of water-soluble polysaccharide component.

    Science.gov (United States)

    Anjaneyalu, Y V; Jagadish, R L; Raju, T S

    1997-06-01

    Polysaccharide components present in the pseudo-stem (scape) of M. paradisiaca were purified from acetone powder of the scape by delignification followed by extraction with aqueous solvents into water soluble polysaccharide (WSP), EDTA-soluble polysaccharide (EDTA-SP), alkali-soluble polysaccharide (ASP) and alkali-insoluble polysaccharide (AISP) fractions. Sugar compositional analysis showed that WSP and EDTA-SP contained only D-Glc whereas ASP contained D-Glc, L-Ara and D-Xyl in approximately 1:1:10 ratio, respectively, and AISP contained D-Glc, L-Ara and D-Xyl in approximately 10:1:2 ratio, respectively. WSP was further purified by complexation with iso-amylalcohol and characterized by specific rotation, IR spectroscopy, Iodine affinity, ferricyanide number, blue value, hydrolysis with alpha-amylase and glucoamylase, and methylation linkage analysis, and shown to be a amylopectin type alpha-D-glucan.

  7. Electrical and non-electrical environment of wind turbine main components

    DEFF Research Database (Denmark)

    Holboell, J.; Henriksen, M.; Olsen, R.S.

    of the electrical components or even lead to catastrophic component failure. In the present paper, results are presented from investigations on existing standards which give detailed descriptions of the environmental and operational conditions of wind turbine components. It is found that there is currently a lack...... of application standards for wind turbine electrical equipment. Component-level environmental requirements as given in equipment-specific standards are compared with the environment described in the IEC's 61400 series concerning wind turbines. Based on methods defined in IEC 60721, the non-electrical environment...... of wind turbine is described by means of specific classes. In the paper, new class combinations are suggested covering the different operating conditions the components are exposed to. The class combinations include factors of climatic, mechanical and chemical character. The factors occur in different...

  8. Company Slogan and a Vivid Image of a Trademark - as Main Components of a Strong Brand

    Directory of Open Access Journals (Sweden)

    Igor BELOSTECINIC

    2015-12-01

    Full Text Available Among other methods, good brand managers always affect clients and strengthen their fidelity to the brand by means of company advertising slogans and vivid images associated with this brand. Slogans may function as useful “levers” or “hooks” helping clients to understand the essence of a trademark and its peculiarity, as well as to nail in clients' minds the established image of a strong brand. One of the “hooks” attracting attention to brand and helping to nail it in a common man's mind – is a company (advertising slogan - short message or motto, which actively presents the key theme of the brand in general or of a certain advertising campaign, and contains substantial information, reflecting the essence of campaign's or product's marketing offer. One must admit that it is nice to recall favorite philosopher's aphorisms or sayings while getting up in the morning, and to whistle an aria from a renowned opera on your way to work. However, more often people sing unsophisticated verses about the “sign of a good taste” and get annoyed by a trivial call to “have a break” going round in their heads. This means that the advertisement of popular chocolate bars and drinks has hit its target. The main lever for success in generation of a positive and recognizable brand image - is the slogan and its graphic image in consumer's mind. Together with other brand components (logo, corporate colors, sound or music image a slogan generates the system of constant elements which ensure the brand's internal unity and are aimed at creation of exposure effect. To some extent, we all are “ad eaters” and are always capable of estimating an ad, even if it is at the level of feelings: whether we like it or not. So why a certain advertisement can make us smile, elicit good feelings and cause an urge to buy product/service that is being advertised, and another one - evokes disappointment and annoyance? The answer is obvious: the first one was

  9. The MainSTREAM Component Platform: A Holistic Approach to Microfluidic System Design

    DEFF Research Database (Denmark)

    Sabourin, David; Skafte-Pedersen, Peder; Søe, Martin Jensen

    2013-01-01

    A microfluidic component library for building systems driving parallel or serial microfluidic-based assays is presented. The components are a miniaturized eight-channel peristaltic pump, an eight-channel valve, sample-to-waste liquid management, and interconnections. The library of components...... of reaction chips; (2) highly parallel pumping and routing/valving capability; (3) methods to interface pumps and chip-to-liquid management systems; (4) means to construct a portable system; (5) reconfigurability/flexibility in system design; (6) means to interface to microscopes; and (7) compatibility...

  10. Fault diagnosis of main coolant pump in the nuclear power station based on the principal component analysis

    International Nuclear Information System (INIS)

    Feng Junting; Xu Mi; Wang Guizeng

    2003-01-01

    The fault diagnosis method based on principal component analysis is studied. The fault character direction storeroom of fifteen parameters abnormity is built in the simulation for the main coolant pump of nuclear power station. The measuring data are analyzed, and the results show that it is feasible for the fault diagnosis system of main coolant pump in the nuclear power station

  11. Procedures for the design of the main mechanical components of a wind system; Dimensionamento dos componentes mecanicos principais de aerogeradores

    Energy Technology Data Exchange (ETDEWEB)

    Hirata, M.H.; Marco Filho, F. de [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia

    1990-12-31

    Procedures for the design of the main mechanical components of a wind system were developed. One of the main concerns was related to the possibility of its use in small micro-computers. This goal was reached and an APPLE II computer was used. The resulting algorithm permits a friendly interaction between man and machine. 5 refs., 12 figs

  12. Remote maintenance development for ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Shibanuma, Kiyoshi

    1997-01-01

    This paper both describes the overall design concept of the ITER remote maintenance system, which has been developed mainly for use with in-vessel components such as divertor and blanket, and outlines of the ITER R and D program, which has been established to develop remote handling equipment/tools and radiation hard components. In ITER, the reactor structures inside cryostat have to be maintained remotely because of activation due to DT operation. Therefore, remote-handling technology is fundamental, and the reactor-structure design must be made consistent with remote maintainability. The overall maintenance scenario and design concepts of the required remote handling equipment/tools have been developed according to their maintenance classification. Technologies are also being developed to verify the feasibility of the maintenance design and include fabrication and testing of a fullscale remote-handling equipment/tools for in-vessel maintenance. (author)

  13. The ITER reduced cost design

    International Nuclear Information System (INIS)

    Aymar, R.

    2000-01-01

    Six years of joint work under the international thermonuclear experimental reactor (ITER) EDA agreement yielded a mature design for ITER which met the objectives set for it (ITER final design report (FDR)), together with a corpus of scientific and technological data, large/full scale models or prototypes of key components/systems and progress in understanding which both validated the specific design and are generally applicable to a next step, reactor-oriented tokamak on the road to the development of fusion as an energy source. In response to requests from the parties to explore the scope for addressing ITER's programmatic objective at reduced cost, the study of options for cost reduction has been the main feature of ITER work since summer 1998, using the advances in physics and technology databases, understandings, and tools arising out of the ITER collaboration to date. A joint concept improvement task force drawn from the joint central team and home teams has overseen and co-ordinated studies of the key issues in physics and technology which control the possibility of reducing the overall investment and simultaneously achieving the required objectives. The aim of this task force is to achieve common understandings of these issues and their consequences so as to inform and to influence the best cost-benefit choice, which will attract consensus between the ITER partners. A report to be submitted to the parties by the end of 1999 will present key elements of a specific design of minimum capital investment, with a target cost saving of about 50% the cost of the ITER FDR design, and a restricted number of design variants. Outline conclusions from the work of the task force are presented in terms of physics, operations, and design of the main tokamak systems. Possible implications for the way forward are discussed

  14. ITER...ation

    International Nuclear Information System (INIS)

    Troyon, F.

    1997-01-01

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

  15. Expected energy fluxes onto ITER Plasma Facing Components during disruption thermal quenches from multi-machine data comparisons

    International Nuclear Information System (INIS)

    Loarte, A.; Andrew, P.; Matthews, G.F.; Paley, J.; Riccardo, V.; Counsell, G.; Eich, T.; Fuchs, C.; Gruber, O.; Herrmann, A.; Pautasso, G.; Federici, G.; Finken, K.H.; Maddaluno, G.; Whyte, D.

    2005-01-01

    A comparison of the power flux characteristics during the thermal quench of plasma disruptions among various tokamak experiments has been carried out and conclusions for ITER have been drawn. It is generally observed that the energy of the plasma at the thermal quench is much smaller than that of a full performance plasma. The timescales for power fluxes onto PFCs during the thermal quench, as determined by IR measurements, are found to scale with device size but not to correlate with pre-disruptive plasma characteristics. The profiles of the thermal quench power fluxes are very broad for diverted discharges, typically a factor of 5-10 broader than that measured during 'normal' plasma operation, while for limiter discharges this broadening is absent. The combination of all the above factors is used to derive the expected range of power fluxes on the ITER divertor target during the thermal quench. The new extrapolation derived in this paper indicates that the average disruption in ITER will deposit an energy flux approximately one order of magnitude lower than previously thought. The evaluation of the ITER divertor lifetime with these revised specifications is carried out. (author)

  16. Identification of the main thermal characteristics of building components using MATLAB

    DEFF Research Database (Denmark)

    Jimenez, M.J.; Madsen, Henrik; Andersen, Klaus Kaae

    2008-01-01

    This paper presents the application of the IDENT Graphical User Interface of MATLAB to estimate the thermal properties of building components from outdoor dynamic testing, imposing appropriate physical constraints and assuming linear and time invariant parametric models. The theory is briefly...

  17. ITER Safety and Licensing

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  18. Biocidal effects of Piper hispidinervum (Piperaceae) essential oil and synergism among its main components.

    Science.gov (United States)

    Andrés, M F; Rossa, G E; Cassel, E; Vargas, R M F; Santana, O; Díaz, C E; González-Coloma, A

    2017-11-01

    In this study we evaluated the effect of a pressure gradient (1-2 atm) in the extraction and composition of the essential oil (EO) of Piper hispidinervum by steam distillation. We also evaluated the insect antifeedant effects (Spodoptera littoralis, Leptinotarsa decemlineata, Myzus persicae and Rhopalosiphum padi) and nematicidal activity (Meloidogyne javanica) of the oils, their major components and their synergistic interactions. Safrole was the major component (78-81%) followed by terpinolene (5-9%). The EOs tested were effective insect antifeedants. Safrole, explained most of the insect antifeedant action of P. hispidinervum EOs. When safrole and terpinolene were tested in binary combinations, low ratios of safrole improved the antifeedant effects of terpinolene. P. hispidinervum EOs caused higher mortality of M. javanica juveniles than their major components. In binary combinations, low ratios of terpinolene increased the nematicidal effects of safrole. The EO treatment strongly suppressed nematode egg hatching and juvenile infectivity. P. hispidinervum EOs affected the germination of S. lycopersicum and L. sativa mostly at 24 h of treatment, being L. sativa the most sensitive. Safrole moderately affected germination and root growth of L. sativa, S. lycopersicum and L. perenne. Terpinolene only affected S. lycopersicum root growth. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. Main features of the pressurized component life management related R and D activity in Hungary

    International Nuclear Information System (INIS)

    Gillemot, F.; Oszwald, F.

    1995-01-01

    During the last one and half years the following main developments related to the life-time management of the NPP units took place in Hungary: 1. National project named AGNES (Advanced General New Evaluation of Safety) finished. 2. The first results of the extended surveillance program of NPP Paks has been measured. 3. The upgraded ultrasonic testing equipment used for RPV outside testing, and nozzle testing is regularly used in refuelling periods. 4. Radiation embrittlement and thermal ageing research started on RPV materials, mainly on cladding. 5. Participation in the development of the IAEA RPV ageing database. 6. Participation in IAEA pilot studies on ageing. 7. Operation of the Budapest Research Reactor, and building a new high capacity helium cooled irradiation rig. 8. Nondestructive evaluation of material ageing of a steam generator vessels. 9. A life-time calculation of Paks RPV-s. This report is a short survey of these developments. 19 refs, 4 figs

  20. A 2 MW, 170 GHz coaxial cavity gyrotron - experimental verification of the design of main components

    Energy Technology Data Exchange (ETDEWEB)

    Piosczyk, B [Forschungszentrum Karlsruhe, Association EURATOM-FZK, Institut fuer Hochleistungsimpuls- und Mikrowellentechnik (IHM), Postfach 3640, D-76021 Karlsruhe (Germany); Dammertz, G [Forschungszentrum Karlsruhe, Association EURATOM-FZK, Institut fuer Hochleistungsimpuls- und Mikrowellentechnik (IHM), Postfach 3640, D-76021 Karlsruhe (Germany); Dumbrajs, O [Department of Engineering Physics and Mathematics, Helsinki University of Technology, Association EURATOM-TEKES, FIN-02150 Espoo (Finland)] (and others)

    2005-01-01

    A 2 MW, CW, 170 GHz coaxial cavity gyrotron is under development in cooperation between European Research Institutions (FZK Karlsruhe, CRPP Lausanne, HUT Helsinki) and the European tube industry (TED, Velizy, France). The design of critical components has recently been examined experimentally at FZK Karlsruhe with a short pulse ({approx} few ms) coaxial cavity gyrotron. This gyrotron uses the same cavity and the same quasioptical (q.o.) RF-output system as designed for the industrial prototype and a very similar electron gun.

  1. Strategic cost management as the main component of strategic management accounting

    OpenAIRE

    Ходзицька, Валентина Василівна

    2013-01-01

    The influence of cost management on making management decisions and functioning of the system of strategic management accounting was analyzed in the paper. The main aspects of the influence of strategic management accounting on making effective management decisions in the system of integrated management of business entities were highlighted. The scope of the organizational activity, covered by the strategic management accounting was described.The paper shows the orientation of strategic manag...

  2. [Productivity costs of rheumatoid arthritis in Germany. Cost composition and prediction of main cost components].

    Science.gov (United States)

    Merkesdal, S; Huelsemann, J L; Mittendorf, T; Zeh, S; Zeidler, H; Ruof, J

    2006-10-01

    Identification of predictors for the productivity cost components: (1) sick leave, and (2) work disability in gainfully employed and (3) impaired household productivity in unemployed patients with rheumatoid arthritis (RA) from the societal perspective. Investigation of productivity costs was linked to a multicenter, randomized, controlled trial evaluating the effectiveness of clinical quality management in 338 patients with RA. The productivity losses were assessed according to the German Guidelines on Health Economic Evaluation. By means of multivariate logistic regression analyses, predictors of sick leave, work disability (employed patients, n=96), and for days confined to bed in unemployed patient (n=242) were determined. Mean annual costs of 970 EUR arose per person taking into consideration all patients (453 EUR sick leave, 63 EUR work disability, 454 EUR impaired productivity of unemployed patients). Disease activity, disease severity, and impaired physical function were global predictors for all of the cost components investigated. Sick leave costs were predicted by prior sick leave periods and the vocational status blue collar worker, work disability costs by sociodemographic variables (marital status, schooling), and the productivity costs of unemployed patients by impaired mental health and impaired physical functions. Interventions such as reduction in disease progression and control of disease activity, early vocational rehabilitation measures and vocational retraining in patients at risk of quitting working life, and self-management programs to learn coping strategies might decrease future RA-related productivity costs.

  3. Diagnosis of the Main Busbar II Panel Components Ageing of RSG-GAS Electrical System by Using Infrared Thermography

    International Nuclear Information System (INIS)

    Teguh Sulistyo; Kiswanto; Roziq Himawan; Ari Satmoko

    2007-01-01

    To support the operation of RSG-GAS safely, the diagnosis of the ageing of main busbar II BHD/BHE/BHF panel components of RSG-GAS electrical system have been done. By using infrared thermography type Thermo Tracer TH9100PM VI/PW VI. The results of the diagnosis showed that some of the components under degradation with various rate. It can cause the system failure. By understanding the components ageing degradation mechanism and performing the preventive and predictive maintenance and safety of RSG-GAS electrical system earlier, the possibility of accident can be avoided. (author)

  4. Status of ITER

    International Nuclear Information System (INIS)

    Aymar, R.

    2002-01-01

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

  5. Identifying and quantifying main components of physiological noise in functional near infrared spectroscopy on prefrontal cortex

    Directory of Open Access Journals (Sweden)

    Evgeniya eKirilina

    2013-12-01

    Full Text Available Functional Near-Infrared Spectroscopy (fNIRS is a promising method to study functional organization of the prefrontal cortex. However, in order to realize the high potential of fNIRS, effective discrimination between physiological noise originating from forehead skin haemodynamic and cerebral signals is required. Main sources of physiological noise are global and local blood flow regulation processes on multiple time scales. The goal of the present study was to identify the main physiological noise contributions in fNIRS forehead signals and to develop a method for physiological de-noising of fNIRS data. To achieve this goal we combined concurrent time-domain fNIRS and peripheral physiology recordings with wavelet coherence analysis. Depth selectivity was achieved by analyzing moments of photon time-of-flight distributions provided by time-domain fNIRS. Simultaneously, mean arterial blood pressure (MAP, heart rate (HR, and skin blood flow (SBF on the forehead were recorded. Wavelet coherence analysis was employed to quantify the impact of physiological processes on fNIRS signals separately for different time scales. We identified three main processes contributing to physiological noise in fNIRS signals on the forehead. The first process with the period of about 3 s is induced by respiration. The second process is highly correlated with time lagged MAP and HR fluctuations with a period of about 10 s often referred as Mayer waves. The third process is local regulation of the facial skin blood flow time locked to the task-evoked fNIRS signals. All processes affect oxygenated haemoglobin concentration more strongly than that of deoxygenated haemoglobin. Based on these results we developed a set of physiological regressors, which were used for physiological de-noising of fNIRS signals. Our results demonstrate that proposed de-noising method can significantly improve the sensitivity of fNIRS to cerebral signals.

  6. A Main Ring bunch length monitor by detecting two frequency components of the beam

    International Nuclear Information System (INIS)

    Ieiri, T.; Jackson, G.

    1989-01-01

    The bunch length is measured by detecting two revolution frequency harmonics of the beam and taking the ratio of their amplitudes. Two heterodyne receivers have been made to direct them, one at 53MHz and the other at 159MHz. These signals are picked-up by a stripline detector. An analog circuit provides a signal proportional to the bunch length. The monitor measures variation of the bunch length as a function of time in the Main Ring. The measured signal, which sometimes shows that the bunches are tumbling in phase space, can be damped by feedback to the RF amplitude modulator. 9 refs., 12 figs., 1 tab

  7. Main error sources in sorbtion technique and plasma electron component parameter definition by continuous X radiation

    International Nuclear Information System (INIS)

    Gavrilov, V.V.; Torokhova, N.V.; Fasakhov, I.K.

    1986-01-01

    Recombination radiation effect on the relation of signals behind the filters depending on the plasma temperature(sorption method for T determination) is demonstrated. This factor produces the main effect on the method accuracy (100-400%), the other factors analysed in combination make an error in temperature at the level of 50%. Method of plasma electron distribution function reconstruction by continuous x-radiation spectrum, based on the correctness (under certain limitations for the required function) of the equation, linking the electron distribution function with bremmsstrahlung spectral density is presented

  8. THE ECONOMETRICS OF THE FORECASTING OF FINANCIAL RESOURCES, A MAIN COMPONENT OF THE FINANCIAL MANAGEMENT

    Directory of Open Access Journals (Sweden)

    2009-05-01

    Full Text Available The paper intends to emphasise the importance of budget resources forecasting for long periods of time, within thefinancial management. An as accurate as possible forecasting of the volume of financial resources will represent the basis forthe future projections of the expenditure of local communities, as they are regulated by law, knowing that one of the principlesrepresenting the basis of budget making is that of the balanced budget. To the same extent, the volume of the budget liquiditieswill depend on the rigorousness of the design of the volume of financial resources.. Beyond the abstract character of themathematic calculus made by specialists in econometrics, the financial manager is also interested to know the forecastingtechniques so that he/she can draw up the income and expenditure budget, the basis for the implementation of the economicsocialdevelopment strategies of the local communities. The financial management remains a fundamental component of thepublic management through the theoretical-methodological arsenal made available for the loan officer.

  9. [Main Components of Xinjiang Lavender Essential Oil Determined by Partial Least Squares and Near Infrared Spectroscopy].

    Science.gov (United States)

    Liao, Xiang; Wang, Qing; Fu, Ji-hong; Tang, Jun

    2015-09-01

    components were 8 in the model. The performance of the model was evaluated according to root mean square error of cross-validation (RMSECV), root mean square error of prediction (RMSEP). In the model, RESECV of linalool and linalyl acetate were 0.170 and 0.416, respectively; RM-SEP were 0.188 and 0.364. The results indicated that raw data was pretreated by OSC and FiPLS, the NIR-PLS quantitative analysis model with good robustness, high measurement precision; it could quickly determine the content of linalool and linalyl acetate in lavender essential oil. In addition, the model has a favorable prediction ability. The study also provide a new effective method which could rapid quantitative analysis the major components of Xinjiang lavender essential oil.

  10. Structural mechanics research and development for main components of chinese 300 MWe PWR NPPs: from design to life management

    International Nuclear Information System (INIS)

    Yao Weida; Dou Yikang; Xie Yongcheng; He Yinbiao; Zhang Ming; Liang Xingyun

    2005-01-01

    Qinshan Nuclear Power Plant (Unit I), is a 300 MWe prototype PWR independently developed by Chinese own efforts, from design, manufacture, construction, installation, commissioning, to operation, inspection, maintenance, ageing management and lifetime assessment. Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has taken up with and involved in deeply the R and D to tackle problems of this type of reactor since very beginning in early 1970s. Structural mechanics is one of the important aspects to ensure the safety and reliability for NPP components. This paper makes a summary on role of structural mechanics for component safety and reliability assessment in different stages of design, commissioning, operation, as well as lifetime assessment on this type PWR NPPs, including Qinshan-I and Chashma-I, a sister plant in Pakistan designed by SNERDI. The main contents of the paper cover design by analysis for key components of NSSS; mechanical problems relating to safety analysis; special problems relating to pressure retaining components, such as fracture mechanics, sealing analysis and its test verifications, etc.; experimental research on flow-induced vibration; seismic qualification for components; component failure diagnosis and root cause analysis; vibration qualification and diagnosis technique; component online monitoring technique; development of defect assessment; methodology of aging management and lifetime assessment for key components of NPPs, etc. (authors)

  11. Product policy - the main component of the marketing mix in the Romanian health services.

    Science.gov (United States)

    Coculescu, B I; Purcarea, V L; Coculescu, E C

    2016-01-01

    The objectives of the reforms in the EU healthcare systems are based on the implementation of the marketing concept in the health systems, which are, among other things: • efficient management of the financial resources and control costs of the rendered health services; • increased satisfaction of the clients of health care services; • broad accessibility to health services; • effective implementation of modern technologies; • rational stimulation of medical services consumption; • achievement of a fair and neutral competition between the public/ private providers and health insurance companies; • introduction of performance criteria in order to increase the incomes of the medical staff and hierarchy in hospitals; • implementation of modern management methods in health services management; • decentralization of the public healthcare system. Product policy in the medical system of healthcare - the most important component of the marketing mix - is the attitude that addresses a medical organization to the volume, structure, and diversity of services subject to their own activities in relation to the requirements of the services market and the competitive actions of other medical institutions.

  12. Antiamoebic Activity of Petiveria alliacea Leaves and Their Main Component, Isoarborinol.

    Science.gov (United States)

    Zavala Ocampo, Lizeth M; Aguirre-HernÁndez, Eva; Pérez-HernÁndez, Nury; Rivera, Gildardo; Marchat, Laurence A; Ramírez-Moreno, Esther

    2017-08-28

    Petiveria alliacea L. (Phytolacaceae) is a medicinal plant with a broad range of traditional therapeutic properties, including the treatment of dysentery and intestinal infections caused by protozoan parasites. However, its effects against Entamoeba histolytica have not been reported yet. We investigated the antiamoebic activity present in the leaves of P. alliacea Antiamoebic activity was evaluated in methanolic and aqueous extracts, as well as in the hexanic, methanolic, and EtOAc fractions. The P. alliacea methanolic extract showed a better antiamoebic activity than the aqueous extract with an IC 50 = 0.51 mg/ml. Likewise, the hexanic fraction was the most effective fraction, showing a dose-dependent activity against E. histolytica , with an IC 50 = 0.68 mg/ml. Hexanic subfraction 12-19 showed the highest antiamoebic activity at 0.8 mg/ml, producing 74.3% growth inhibition without any toxicity in mammal cells. A major component in subfraction 12-19 was identified as isoarborinol, which produced 51.4% E. histolytica growth inhibition at 0.05 mg/ml without affecting mammal cells. The P. alliacea leaf extract has antiamoebic activity that can be attributed to a major metabolite known as isoarborinol.

  13. Tungsten: An option for divertor and main chamber plasma facing components in future fusion devices

    International Nuclear Information System (INIS)

    Neu, R.; Dux, R.; Kallenbach, A.; Maggi, C.F.; Puetterich, T.; Balden, M.; Eich, T.; Fuchs, J.C.; Gruber, O.; Herrmann, A.; Maier, H.; Mueller, H.W.; Pugno, R.; Radivojevic, I.; Rohde, V.; Sips, A.C.C.; Suttrop, W.; Ye, M.Y.; O'Mullane, M.; Whiteford, A.

    2005-01-01

    The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic and the cooling factor of W have been extended and refined. The W-coated surfaces represent now a fraction of 65% (24.8 m2). The only two major components which are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. A very successful remedy is the addition of central RF heating at the 20-30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10 -3 and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper, W coated divertor do not show higher W concentrations than comparable discharges in the lower C-based divertor. (author)

  14. An investigation of pulsed phase thermography for detection of disbonds in HIP-bonded beryllium tiles in ITER normal heat flux first wall (NHF FW) components

    Energy Technology Data Exchange (ETDEWEB)

    Bushell, J., E-mail: joe.bushell@amec.com [AMEC Foster Wheeler, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England (United Kingdom); Sherlock, P. [AMEC Foster Wheeler, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England (United Kingdom); Mummery, P. [School of Mechanical, Aerospace and Civil Engineering, University of Manchester, England (United Kingdom); Bellin, B.; Zacchia, F. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, Barcelona (Spain)

    2015-10-15

    Highlights: • Pulsed phase thermography was trialled on Be-tiled plasma facing components. • Two components, one with known disbonds, one intact, were inspected and compared. • Finite element analysis was used to verify experimental observations. • PPT successfully detected disbonds in the failed component. • Good agreement found with ultrasonic test, though defect geometry was uncertain. - Abstract: Pulsed phase thermography (PPT) is a non destructive examination (NDE) technique, traditionally used in the Aerospace Industry for inspection of composite structures, which combines characteristics and benefits of flash thermography and lock-in thermography into a single, rapid inspection technique. The aim of this work was to evaluate the effectiveness of PPT as a means of inspection for the bond between the beryllium (Be) tiles and the copper alloy (CuCrZr) heatsink of the ITER NHF FW components. This is a critical area dictating the functional integrity of these components, as single tile detachment in service could result in cascade failure. PPT has advantages over existing thermography techniques using heated water which stress the component, and the non-invasive, non-contact nature presents advantages over existing ultrasonic methods. The rapid and non-contact nature of PPT also gives potential for in-service inspections as well as a quality measure for as-manufactured components. The technique has been appraised via experimental trials using ITER first wall mockups with pre-existing disbonds confirmed via ultrasonic tests, partnered with finite element simulations to verify experimental observations. This paper will present the results of the investigation.

  15. An investigation of pulsed phase thermography for detection of disbonds in HIP-bonded beryllium tiles in ITER normal heat flux first wall (NHF FW) components

    International Nuclear Information System (INIS)

    Bushell, J.; Sherlock, P.; Mummery, P.; Bellin, B.; Zacchia, F.

    2015-01-01

    Highlights: • Pulsed phase thermography was trialled on Be-tiled plasma facing components. • Two components, one with known disbonds, one intact, were inspected and compared. • Finite element analysis was used to verify experimental observations. • PPT successfully detected disbonds in the failed component. • Good agreement found with ultrasonic test, though defect geometry was uncertain. - Abstract: Pulsed phase thermography (PPT) is a non destructive examination (NDE) technique, traditionally used in the Aerospace Industry for inspection of composite structures, which combines characteristics and benefits of flash thermography and lock-in thermography into a single, rapid inspection technique. The aim of this work was to evaluate the effectiveness of PPT as a means of inspection for the bond between the beryllium (Be) tiles and the copper alloy (CuCrZr) heatsink of the ITER NHF FW components. This is a critical area dictating the functional integrity of these components, as single tile detachment in service could result in cascade failure. PPT has advantages over existing thermography techniques using heated water which stress the component, and the non-invasive, non-contact nature presents advantages over existing ultrasonic methods. The rapid and non-contact nature of PPT also gives potential for in-service inspections as well as a quality measure for as-manufactured components. The technique has been appraised via experimental trials using ITER first wall mockups with pre-existing disbonds confirmed via ultrasonic tests, partnered with finite element simulations to verify experimental observations. This paper will present the results of the investigation.

  16. 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.

  17. A Monte Carlo study of Weibull reliability analysis for space shuttle main engine components

    Science.gov (United States)

    Abernethy, K.

    1986-01-01

    The incorporation of a number of additional capabilities into an existing Weibull analysis computer program and the results of Monte Carlo computer simulation study to evaluate the usefulness of the Weibull methods using samples with a very small number of failures and extensive censoring are discussed. Since the censoring mechanism inherent in the Space Shuttle Main Engine (SSME) data is hard to analyze, it was decided to use a random censoring model, generating censoring times from a uniform probability distribution. Some of the statistical techniques and computer programs that are used in the SSME Weibull analysis are described. The methods documented in were supplemented by adding computer calculations of approximate (using iteractive methods) confidence intervals for several parameters of interest. These calculations are based on a likelihood ratio statistic which is asymptotically a chisquared statistic with one degree of freedom. The assumptions built into the computer simulations are described. The simulation program and the techniques used in it are described there also. Simulation results are tabulated for various combinations of Weibull shape parameters and the numbers of failures in the samples.

  18. Trend and pattern analysis of failures of main feedwater system components in United States commercial nuclear power plants

    International Nuclear Information System (INIS)

    Gentillon, C.D.; Meachum, T.R.; Brady, B.M.

    1987-01-01

    The goal of the trend and pattern analysis of MFW (main feedwater) component failure data is to identify component attributes that are associated with relatively high incidences of failure. Manufacturer, valve type, and pump rotational speed are examples of component attributes under study; in addition, the pattern of failures among NPP units is studied. A series of statistical methods is applied to identify trends and patterns in failures and trends in occurrences in time with regard to these component attributes or variables. This process is followed by an engineering evaluation of the statistical results. In the remainder of this paper, the characteristics of the NPRDS that facilitate its use in reliability and risk studies are highlighted, the analysis methods are briefly described, and the lessons learned thus far for improving MFW system availability and reliability are summarized (orig./GL)

  19. Experimental study of divertor plasma-facing components damage under a combination of pulsed and quasi-stationary heat loads relevant to expected transient events at ITER

    International Nuclear Information System (INIS)

    Klimov, N S; Podkovyrov, V L; Kovalenko, D V; Zhitlukhin, A M; Barsuk, V A; Mazul, I V; Giniyatulin, R N; Kuznetsov, V Ye; Riccardi, B; Loarte, A; Merola, M; Koidan, V S; Linke, J; Landman, I S; Pestchanyi, S E; Bazylev, B N

    2011-01-01

    This paper concerns the experimental study of damage of ITER divertor plasma-facing components (PFCs) under a combination of pulsed plasma heat loads (representative of controlled ITER type I edge-localized modes (ELMs)) and quasi-stationary heat loads (representative of the high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events). The PFC's tungsten armor damage under pulsed plasma exposure was driven by (i) the melt layer motion, which leads to bridges formation between neighboring tiles and (ii) the W brittle failure giving rise to a stable crack pattern on the exposed surface. The crack width reaches a saturation value that does not exceed some tens of micrometers after several hundreds of ELM-like pulses. HHF thermal fatigue tests have shown (i) a peeling-off of the re-solidified material due to its brittle failure and (ii) a significant widening (up to 10 times) of the cracks and the formation of additional cracks.

  20. Screening for main components associated with the idiosyncratic hepatotoxicity of a tonic herb, Polygonum multiflorum.

    Science.gov (United States)

    Li, Chunyu; Niu, Ming; Bai, Zhaofang; Zhang, Congen; Zhao, Yanling; Li, Ruiyu; Tu, Can; Li, Huifang; Jing, Jing; Meng, Yakun; Ma, Zhijie; Feng, Wuwen; Tang, Jinfa; Zhu, Yun; Li, Jinjie; Shang, Xiaoya; Zou, Zhengsheng; Xiao, Xiaohe; Wang, Jiabo

    2017-06-01

    The main constituents of a typical medicinal herb, Polygonum multiflorum (Heshouwu in Chinese), that induces idiosyncratic liver injury remain unclear. Our previous work has shown that cotreatment with a nontoxic dose of lipopolysaccharide (LPS) and therapeutic dose of Heshouwu can induce liver injury in rats, whereas the solo treatment cannot induce observable injury. In the present work, using the constituent "knock-out" and "knock-in" strategy, we found that the ethyl acetate (EA) extract of Heshouwu displayed comparable idiosyncratic hepatotoxicity to the whole extract in LPS-treated rats. Results indicated a significant elevation of plasma alanine aminotransferase, aspartate aminotransferase, and liver histologic changes, whereas other separated fractions failed to induce liver injury. The mixture of EA extract with other separated fractions induced comparable idiosyncratic hepatotoxicity to the whole extract in LPS-treated rats. Chemical analysis further revealed that 2,3,5,4'-tetrahydroxy trans-stilbene-2-O-β-glucoside (trans-SG) and its cis-isomer were the two major compounds in EA extract. Furthermore, the isolated cis-, and not its trans-isomer, displayed comparable idiosyncratic hepatotoxicity to EA extract in LPS-treated rats. Higher contents of cis-SG were detected in Heshouwu liquor or preparations from actual liver intoxication patients associated with Heshouwu compared with general collected samples. In addition, plasma metabolomics analysis showed that cis-SG-disturbing enriched pathways remarkably differed from trans-SG ones in LPS-treated rats. All these results suggested that cis-SG was closely associated with the idiosyncratic hepatotoxicity of Heshouwu. Considering that the cis-trans isomerization of trans-SG was mediated by ultraviolet light or sunlight, our findings serve as reference for controlling photoisomerization in drug discovery and for the clinical use of Heshouwu and stilbene-related medications.

  1. Design improvements and R and D achievements for VV and in-vessel components towards ITER construction and implications for the R and D programme

    International Nuclear Information System (INIS)

    Ioki, K.

    2002-01-01

    Procurement specifications are now being finalised for ITER components whose construction is lengthy, yet which are needed early, such as the vacuum vessel. Although the basic concept of the vacuum vessel (VV) and in-vessel components of the ITER design has stayed the same as reported at the last conference, there have been several detailed design improvements resulting from efforts to raise reliability, to improve better maintainability and to save money. One of the most important achievements in the VV R and D is demonstration of the necessary assembly tolerances. Further development of advanced methods of cutting, welding and NDT for the VV have been continued in order to refine manufacturing and improve cost and technical performance. With regard to the related FW/blanket and divertor designs, the R and D has resulted in the development of suitable technologies. Prototypes of the FW panel, the blanket shield block and the divertor components have been successfully fabricated. This paper reviews the recent progress in the design as procurement nears. (author)

  2. Analysis of the main components of the aguamiel produced by the maguey-pulquero (Agave mapisaga) throughout the harvest period.

    Science.gov (United States)

    Ortiz-Basurto, Rosa Isela; Pourcelly, Gérald; Doco, Thierry; Williams, Pascale; Dornier, Manuel; Belleville, Marie-Pierre

    2008-05-28

    The main characteristics of the aguamiel (maguey-pulquero sap) during the harvest period of the Agave mapisaga plants were assessed to establish its stability through time and the industrial potential of its components. Only minor differences in aguamiel composition were detected among samples collected at different time points of the harvest period. The aguamiel analyzed contained 11.5 wt % of dry matter, which was composed mainly of sugars (75 wt %). Among these sugars, 10 wt % were fructo-oligosaccharides (FOS), which are known to be important in the food industry for their prebiotic properties. Other components include 0.3 wt % of free amino acids (with most essential amino acids and four neurotransmitters: GABA, GLY, GLX, and ASX), 3 wt % of proteins, and 3 wt % of ashes.

  3. The permeability characteristics and interaction of main components from Si-Ni-San in a MDCK epithelial cell monolayer model.

    Science.gov (United States)

    Chen, Ruonan; Shen, Chenlin; Xu, Qingqing; Liu, Yaru; Li, Bo; Huang, Cheng; Ma, Taotao; Meng, Xiaoming; Wu, Maomao; Li, Jun

    2017-07-26

    1. Si-Ni-San (SNS) possesses extensive therapeutic effects, however, the extent to which main components are absorbed and the mechanisms involved are controversial. 2. In this study, MDCK cell model was used to determine the permeability characteristics and interaction between the major components of Si-Ni-San, including saikosaponin a, paeoniflorin, naringin and glycyrrhizic acid. 3. The transport of the major components was concentration-dependent in both directions. Moreover, the transport of paeoniflorin, naringin and glycyrrhizic acid was significantly reduced at 4°C or in the presence of NaN3. Additionally, the efflux of paeoniflorin and naringin were apparently reduced in the presence of P-gp inhibitor verapamil. The transport of glycyrrhizic acid was clearly inhibited by the inhibitors of MRP2, indicating that MRP2 may be involved in the transport of glycyrrhizic acid. However, the results indicated that saikosaponin a was absorbed mainly by passive diffusion. Furthermore, the combined incubation of four major components had a powerful sorbefacient effect than a single drug used alone which may be regulated by tight junctions. 4. Taken together, our study provides useful information for pharmacological applications of Si-Ni-San and offers new insights into this ancient decoction for further researches, especially in drug synergism.

  4. ITER ITA newsletter. No. 8, September 2003

    International Nuclear Information System (INIS)

    2003-10-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities including Robert Aymar's leaving ITER for CERN, ITER related issues at the IAEA General Conference and status and prospects of thermonuclear power and activity during the ITA on materials foe vessel and in-vessel components

  5. Plasma position and shape control for ITER

    International Nuclear Information System (INIS)

    Portone, A.; Gribov, Y.; Huguet, M.

    1995-01-01

    Key features and main results about the control of the plasma shape in ITER are presented. A control algorithm is designed to control up to 6 gaps between the plasma separatrix and the plasma facing components during the reference burn phase. Nonlinear simulations show the performances of the controller in the presence of plasma vertical position offsets, beta drops and power supply voltage saturation

  6. Specialized data analysis for the Space Shuttle Main Engine and diagnostic evaluation of advanced propulsion system components

    Science.gov (United States)

    1993-01-01

    The Marshall Space Flight Center is responsible for the development and management of advanced launch vehicle propulsion systems, including the Space Shuttle Main Engine (SSME), which is presently operational, and the Space Transportation Main Engine (STME) under development. The SSME's provide high performance within stringent constraints on size, weight, and reliability. Based on operational experience, continuous design improvement is in progress to enhance system durability and reliability. Specialized data analysis and interpretation is required in support of SSME and advanced propulsion system diagnostic evaluations. Comprehensive evaluation of the dynamic measurements obtained from test and flight operations is necessary to provide timely assessment of the vibrational characteristics indicating the operational status of turbomachinery and other critical engine components. Efficient performance of this effort is critical due to the significant impact of dynamic evaluation results on ground test and launch schedules, and requires direct familiarity with SSME and derivative systems, test data acquisition, and diagnostic software. Detailed analysis and evaluation of dynamic measurements obtained during SSME and advanced system ground test and flight operations was performed including analytical/statistical assessment of component dynamic behavior, and the development and implementation of analytical/statistical models to efficiently define nominal component dynamic characteristics, detect anomalous behavior, and assess machinery operational condition. In addition, the SSME and J-2 data will be applied to develop vibroacoustic environments for advanced propulsion system components, as required. This study will provide timely assessment of engine component operational status, identify probable causes of malfunction, and indicate feasible engineering solutions. This contract will be performed through accomplishment of negotiated task orders.

  7. [Multiple analysis of the difference in intestinal absorption between the main components and the extract of Glycyrrhiza uralensis].

    Science.gov (United States)

    Wu, Qing-Qing; Chen, Yan; Xin, Ran; Wang, Jin-Yan; Zhou, Lei; Yuan, Ling; Jia, Xiao-Bin

    2012-05-01

    The aim of this study is to investigate the rat intestinal absorption behavior of two main active components, liquiritin, glycyrrhizin and the extract of Glycyrrhiza uralensis. The rat intestinal perfusion model was employed. Concentrations of the compounds of the interest in the intestinal perfusate, bile and plasma samples were determined by HPLC and UPLC. At the same time, the intestinal enzymes incubation test and the partition coefficient determination, the absorption of liquiritin and glycyrrhizin alone and the extract were multiple analyzed. The results showed that the P(eff) (effective permeability) of liquiritin or glycyrrhizin alone or the extract was less than 0.3, which suggested their poor absorption in the intestine. The P(eff) of the two main active components or the extract was not significantly different in duodenum, jejunum, colon and ileum segment. The P(eff) of the glycyrrhizin in the extract had no significant difference in the four intestinal segments compared with the glycyrrhizin alone. The absorption of the liquiritin displayed significant difference (P components might not increase the amount of liquiritin and glycyrrhizin in the bile and plasma within the duration of the test.

  8. ITER management advisory committee meeting in NAKA

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    1999-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held on 17 December 1999 in Naka, Japan. The main topics were the ITER EDA Status, Task Status Summary and Work Program and a schedule of ITER meetings

  9. Total annoyance from an industrial noise source with a main spectral component combined with a background noise.

    Science.gov (United States)

    Alayrac, M; Marquis-Favre, C; Viollon, S

    2011-07-01

    When living close to an industrial plant, people are exposed to a combination of industrial noise sources and a background noise composed of all the other noise sources in the environment. As a first step, noise annoyance indicators in laboratory conditions are proposed for a single exposure to an industrial noise source. The second step detailed in this paper involves determining total annoyance indicators in laboratory conditions for ambient noises composed of an industrial noise source and a background noise. Two types of steady and permanent industrial noise sources are studied: low frequency noises with a main spectral component at 100 Hz, and noises with a main spectral component in middle frequencies. Five background noises are assessed so as to take into account different sound environments which can usually be heard by people living around an industrial plant. One main conclusion of this study is that two different analyses are necessary to determine total annoyance indicators for this type of ambient noise, depending on the industrial noise source composing it. Therefore, two total annoyance indicators adapted to the ambient noises studied are proposed. © 2011 Acoustical Society of America

  10. ITER EDA status

    International Nuclear Information System (INIS)

    Aymar, R.

    2001-01-01

    The Project has focused on drafting the Plant Description Document (PDD), which will be published as the Technical Basis for the ITER Final Design Report (FDR), and its related documentation in time for the ITER review process. The preparations have involved continued intensive detailed design work, analyses and assessments by the Home Teams and the Joint Central Team, who have co-operated closely and efficiently. The main technical document has been completed in time for circulation, as planned, to TAC members for their review at TAC-17 (19-22 February 2001). Some of the supporting documents, such as the Plant Design Specification (PDS), Design Requirements and Guidelines (DRG1 and DRG2), and the Plant Safety Requirement (PSR) are also available for reference in draft form. A summary paper of the PDD for the Council's information is available as a separate document. A new documentation structure for the Project has been established. This hierarchical structure for documentation facilitates the entire organization in a way that allows better change control and avoids duplications. The initiative was intended to make this documentation system valid for the construction and operation phases of ITER. As requested, the Director and the JCT have been assisting the Explorations to plan for future joint technical activities during the Negotiations, and to consider technical issues important for ITER construction and operation for their introduction in the draft of a future joint implementation agreement. As charged by the Explorers, the Director has held discussions with the Home Team Leaders in order to prepare for the staffing of the International Team and Participants Teams during the Negotiations (Co-ordinated Technical Activities, CTA) and also in view of informing all ITER staff about their future directions in a timely fashion. One important element of the work was the completion by the Parties' industries of costing studies of about 83 ''procurement packages

  11. ITER: a technology test bed for a fusion reactor

    International Nuclear Information System (INIS)

    Huguet, M.; Green, B.J.

    1996-01-01

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

  12. ITER and the fusion reactor: status and challenge to technology

    International Nuclear Information System (INIS)

    Lackner, K.

    2001-01-01

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

  13. Vimang (Mangifera indica L. extract) induces permeability transition in isolated mitochondria, closely reproducing the effect of mangiferin, Vimang's main component.

    Science.gov (United States)

    Pardo-Andreu, Gilberto L; Dorta, Daniel Junqueira; Delgado, René; Cavalheiro, Renata A; Santos, Antonio C; Vercesi, Anibal E; Curti, Carlos

    2006-02-01

    Mitochondrial permeability transition (MPT) is a Ca(2+)-dependent, cyclosporin A (CsA)-sensitive, non-selective inner membrane permeabilization process. It is often associated with apoptotic cell death, and is induced by a wide range of agents or conditions, usually involving reactive oxygen species (ROS). In this study, we demonstrated that Mangifera indica L. extract (Vimang), in the presence of 20 microM Ca(2+), induces MPT in isolated rat liver mitochondria, assessed as CsA-sensitive mitochondrial swelling, closely reproducing the same effect of mangiferin, the main component of the extract, as well as MPT-linked processes like oxidation of membrane protein thiols, mitochondrial membrane potential dissipation and Ca(2+) release from organelles. The flavonoid catechin, the second main component of Vimang, also induces MPT, although to a lesser extent; the minor, but still representative Vimang extract components, gallic and benzoic acids, show respectively, low and high MPT inducing abilities. Nevertheless, following exposure to H(2)O(2)/horseradish peroxidase, the visible spectra of these compounds does not present the same changes previously reported for mangiferin. It is concluded that Vimang-induced MPT closely reproduces mangiferin effects, and proposed that this xanthone is the main agent responsible for the extract's MPT inducing ability, by the action on mitochondrial membrane protein thiols of products arising as a consequence of the mangiferin's antioxidant activity. While this effect would oppose the beneficial effect of Vimang's antioxidant activity, it could nevertheless benefit cells exposed to over-production of ROS as occurring in cancer cells, in which triggering of MPT-mediated apoptosis may represent an important defense mechanism to their host.

  14. Structural impact of armor monoblock dimensions on the failure behavior of ITER-type divertor target components: Size matters

    Energy Technology Data Exchange (ETDEWEB)

    Li, Muyuan; You, Jeong-Ha, E-mail: you@ipp.mpg.de

    2016-12-15

    Highlights: • Quantitative assessment of size effects was conducted numerically for W monoblock. • Decreasing the width of W monoblock leads to a lower risk of failure. • The Cu interlayer was not affected significantly by varying armor thickness. • The predicted trends were in line with the experimental observations. - Abstract: Plenty of high-heat-flux tests conducted on tungsten monoblock type divertor target mock-ups showed that the threshold heat flux density for cracking and fracture of tungsten armor seems to be related to the dimension of the monoblocks. Thus, quantitative assessment of such size effects is of practical importance for divertor target design. In this paper, a computational study about the thermal and structural impact of monoblock size on the plastic fatigue and fracture behavior of an ITER-type tungsten divertor target is reported. As dimensional parameters, the width and thickness of monoblock, the thickness of sacrificial armor, and the inner diameter of cooling tube were varied. Plastic fatigue lifetime was estimated for the loading surface of tungsten armor and the copper interlayer by use of a cyclic-plastic constitutive model. The driving force of brittle crack growth through the tungsten armor was assessed in terms of J-integral at the crack tip. Decrease of the monoblock width effectively reduced accumulation of plastic strain at the armor surface and the driving force of brittle cracking. Decrease of sacrificial armor thickness led to decrease of plastic deformation at the loading surface due to lower surface temperature, but the thermal and mechanical response of the copper interlayer was not affected by the variation of armor thickness. Monoblock with a smaller tube diameter but with the same armor thickness and shoulder thickness experienced lower fatigue load. The predicted trends were in line with the experimental observations.

  15. Rapid analysis of the main components of the total glycosides of Ranunculus japonicus by UPLC/Q-TOF-MS.

    Science.gov (United States)

    Rui, Wen; Chen, Hongyuan; Tan, Yuzhi; Zhong, Yanmei; Feng, Yifan

    2010-05-01

    A rapid method for the analysis of the main components of the total glycosides of Ranunculus japonicus (TGOR) was developed using ultra-performance liquid chromatography with quadrupole-time-of-flight mass spectrometry (UPLC/Q-TOF-MS). The separation analysis was performed on a Waters Acquity UPLC system and the accurate mass of molecules and their fragment ions were determined by Q-TOF MS. Twenty compounds, including lactone glycosides, flavonoid glycosides and flavonoid aglycones, were identified and tentatively deduced on the basis of their elemental compositions, MS/MS data and relevant literature. The results demonstrated that lactone glycosides and flavonoids were the main constituents of TGOR. Furthermore, an effective and rapid pattern was established allowing for the comprehensive and systematic characterization of the complex samples.

  16. Coptidis rhizoma and its main bioactive components: recent advances in chemical investigation, quality evaluation and pharmacological activity.

    Science.gov (United States)

    Meng, Fan-Cheng; Wu, Zheng-Feng; Yin, Zhi-Qi; Lin, Li-Gen; Wang, Ruibing; Zhang, Qing-Wen

    2018-01-01

    Coptidis rhizoma (CR) is the dried rhizome of Coptis chinensis Franch., C. deltoidea C. Y. Cheng et Hsiao or C. teeta Wall. (Ranunculaceae) and is commonly used in Traditional Chinese Medicine for the treatment of various diseases including bacillary dysentery, typhoid, tuberculosis, epidemic cerebrospinal meningitis, empyrosis, pertussis, and other illnesses. A literature survey was conducted via SciFinder, ScieneDirect, PubMed, Springer, and Wiley databases. A total of 139 selected references were classified on the basis of their research scopes, including chemical investigation, quality evaluation and pharmacological studies. Many types of secondary metabolites including alkaloids, lignans, phenylpropanoids, flavonoids, phenolic compounds, saccharides, and steroids have been isolated from CR. Among them, protoberberine-type alkaloids, such as berberine, palmatine, coptisine, epiberberine, jatrorrhizine, columamine, are the main components of CR. Quantitative determination of these alkaloids is a very important aspect in the quality evaluation of CR. In recent years, with the advances in isolation and detection technologies, many new instruments and methods have been developed for the quantitative and qualitative analysis of the main alkaloids from CR. The quality control of CR has provided safety for pharmacological applications. These quality evaluation methods are also frequently employed to screen the active components from CR. Various investigations have shown that CR and its main alkaloids exhibited many powerful pharmacological effects including anti-inflammatory, anti-cancer, anti-diabetic, neuroprotective, cardioprotective, hypoglycemic, anti-Alzheimer and hepatoprotective activities. This review summarizes the recent phytochemical investigations, quality evaluation methods, the biological studies focusing on CR as well as its main alkaloids.

  17. 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

  18. Reactivity of main components and substituent distribution in esterified sugarcane bagasse prepared by effective solid phase reaction.

    Science.gov (United States)

    Gan, Tao; Zhang, Yanjuan; Chen, Yane; Hu, Huayu; Yang, Mei; Huang, Zuqiang; Chen, Dong; Huang, Aimin

    2018-02-01

    Three main components of lignocellulose (cellulose, hemicellulose, and lignin isolated from sugarcane bagasse (SCB)) as well as holocellulose and SCB were modified with maleic acid by mechanical activation (MA)-assisted solid phase reaction (MASPR) technology. The order of reactivity was found to be lignin>hemicellulose>cellulose. The amorphous structure of lignin and hemicellulose mainly attributed to their better reactivity, and the modified lignin could reach a maximum degree of esterification (DE) of 93.45%. MA improved the accessibility and reactivity of cellulose, as the DE of modified cellulose gradually increased with milling time and reached the maximum value of 57.30% at 120min, which had significant effect on structure changes and DE of modified holocellulose and SCB. MA enhanced the esterification of all three components in lignocellulose with relatively high substituent distribution in them, and maleated SCB with a maximum DE of 64.17% was successfully prepared by this simple, green, and effective MASPR technology. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. [Comparative study on specific chromatograms and main active components of wild and cultivated rhizomes of Paris polyphylla var. yunnanensis].

    Science.gov (United States)

    Zan, Ke; Gao, Yu-Ming; Cui, Gan; Liu, Jie; Guo, Li-Nong; Zheng, Jian; Ma, Shuang-Cheng

    2017-08-01

    The present study is to compare specific chromatograms and main acitive components between wild and cultivated rhizomes of Paris polyphylla var. yunnanensis by HPLC. HPLC analysis was performed on a Waters XSelect HSS T3 C₁₈ clumn (4.6 mm×250 mm, 5 μm), with a mobile phase consisting of acetonitrile (A)-water (B) at a flow rate of 1 mL•min⁻¹ (0-50 min,30%-50%A;50-80 min,50% A,80-85 min,50%-30%A;85-100 min,30% A). The detection wavelength was 203 nm and the column temperature was controlled at 30 ℃, and the injection volume was 10 μL. HPLC specific chromatograms of wild and cultivated rhizomes of P. polyphylla var. yunnanensis were established and nine steroidal saponins were simultaneously determined by the above method. The mean contents of paris saponin Ⅶ, paris saponin H and total average contents of four pennogenyl saponins in Rhizomes of wild samples were significantly higher than those of cultivated ones. However, this result is opposite from the average content of paris saponin Ⅰ and total average contents of five dioscins in the wild and cultivated samples. Because the significant differences occurred for the specific chromatograms and main active components between the wild and cultivated P. polyphylla var. yunnanensis, much more pharmacological and clinical researches are therefore necessary. Copyright© by the Chinese Pharmaceutical Association.

  20. High heat flux components in fusion devices: from nowadays experience in Tore Supra towards the ITER challenge

    International Nuclear Information System (INIS)

    Grosman, A.; Bayetti, P.; Chappuis, P.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Guilhem, D.; Lipa, M.; Marbach, G.; Mitteau, R.; Schlosser, J.

    2003-01-01

    A pioneering activity has been developed by CEA and the European industry in the field of actively cooled high heat flux plasma facing components in Tore Supra operation, which is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW.m -2 of nominal convected heat flux. This success is the result of a long lead development and industrialization program (about 10 years) marked out with a number of technical and managerial challenges that were taken up and has allowed us to build up an unique experience feedback database. This is illustrated in this paper with the specific example of the development of high heat flux CFC-on-CuCrZr (carbon-carbon fibre composite on hardened copper alloy CuCrZr) component from design phase to tokamak operation. (authors)

  1. High heat flux components in fusion devices: from nowadays experience in Tore Supra towards the ITER challenge

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, A.; Bayetti, P.; Chappuis, P.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Guilhem, D.; Lipa, M.; Marbach, G.; Mitteau, R.; Schlosser, J

    2003-07-01

    A pioneering activity has been developed by CEA and the European industry in the field of actively cooled high heat flux plasma facing components in Tore Supra operation, which is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW.m{sup -2} of nominal convected heat flux. This success is the result of a long lead development and industrialization program (about 10 years) marked out with a number of technical and managerial challenges that were taken up and has allowed us to build up an unique experience feedback database. This is illustrated in this paper with the specific example of the development of high heat flux CFC-on-CuCrZr (carbon-carbon fibre composite on hardened copper alloy CuCrZr) component from design phase to tokamak operation. (authors)

  2. Study of the WWR-S IFIN-HH reactor main components stare, after 40 years working, using nondestructive methods

    International Nuclear Information System (INIS)

    Dragolici, A. C.; Zorliu, A.; Ripeanu, R.; Petran, C.; Mincu, I.

    2000-01-01

    The main goal of these investigations was to establish the security level after 40 years of working of the WWR-S research reactor of Horia Hulubei National Institute of Research and Development for Physics and Nuclear Engineering, Bucharest-Magurele. The purpose of these investigations was: checking the functionality and the physical integrity of the main components of the reactor. The physical integrity of the components is usually affected by slow processes, such as: corrosion, erosion, aging, deformations and initially hidden flaws with very slow evolutions. The methods used to determine the effects of these processes and to infer conclusions about the physical integrity of the facility are: visualizations by optical means (endoscopy and video camera), examination using ultrasounds and gammagraphy. The objective of the endoscopic checking was the view of the state of interior surfaces of the tubes and pipes, specially the inaccessible areas of the non-dismantling parts of the reactor. Big size components, such as reactor vessel, the biologic protection vessel and the main large diameter pipes of the primary cooling system, were investigated using a special device that contains a video camera connected to a PC. To obtain more information regarding the evolution of the corrosion spots, scratches and harmed areas on the investigated surfaces, their depth was checked by ultrasounds, and the welding seams structure was determined by gammagraphy. A table is given with some significant results obtained from ultrasound measurements in different points of reactor vessel, thermal column, horizontal tubes, etc. After these tests, the conclusions are: the maximum corrosion depth is 0.2 mm; - scratches are superficially, not exceeding 0.2-0.5 mm; - the traces of harmed areas are produced by the electromagnetic device utilization used for manipulation of aluminium capsules which contain irradiated substances. They are superficial, with maximum area of about 1 cm 2 ; the

  3. Essential oil from Zanthoxylum bungeanum Maxim. and its main components used as transdermal penetration enhancers: a comparative study.

    Science.gov (United States)

    Lan, Yi; Li, Hui; Chen, Yan-yan; Zhang, Ye-wen; Liu, Na; Zhang, Qing; Wu, Qing

    2014-11-01

    Our previous studies had confirmed that the essential oil from Zanthoxylum bungeanum Maxim. (Z. bungeanum oil) could effectively enhance the percutaneous permeation of drug molecules as a natural transdermal penetration enhancer. The aim of the present study is to investigate and compare the skin penetration enhancement effect of Z. bungeanum oil and its main components on traditional Chinese medicine (TCM) active components. Toxicities of Z. bungeanum oil and three selected terpene compounds (terpinen-4-ol, 1,8-cineole, and limonene) in epidermal keratinocytes (HaCaT) and dermal fibroblast (CCC-ESF-1) cell lines were measured using an MTT (3-(4,5-dimethylthiazol-2-yl)-2,5-diphenyltetrazolium bromide) assay. Five model drugs in TCM external preparations, namely osthole (OT), tetramethylpyrazine (TMP), ferulic acid (FA), puerarin (PR), and geniposide (GP), which were selected based on their lipophilicity denoted by logKo/w, were tested using in vitro permeation studies in which vertical Franz diffusion cells and rat abdominal skin were employed. The secondary structure changes of skin stratum corneum (SC) and drug thermodynamic activities were investigated to understand their mechanisms of action using Fourier transform infrared (FTIR) spectroscopy and saturation solubility studies, respectively. It was found that Z. bungeanum oil showed lower toxicities in both HaCaT cells and CCC-ESF-1 cells compared with three terpene compounds used alone. The enhancement permeation capacities by all tested agents were in the following increasing order: terpinen-4-ol≈1,8-cineolemainly by affecting SC lipids. These results indicated that Z. bungeanum oil exhibited better performance in enhancing the skin permeation of active components in TCM preparations.

  4. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

    1993-01-01

    Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

  5. 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.)

  6. Multichannel Signals Reconstruction Based on Tunable Q-Factor Wavelet Transform-Morphological Component Analysis and Sparse Bayesian Iteration for Rotating Machines

    Directory of Open Access Journals (Sweden)

    Qing Li

    2018-04-01

    Full Text Available High-speed remote transmission and large-capacity data storage are difficult issues in signals acquisition of rotating machines condition monitoring. To address these concerns, a novel multichannel signals reconstruction approach based on tunable Q-factor wavelet transform-morphological component analysis (TQWT-MCA and sparse Bayesian iteration algorithm combined with step-impulse dictionary is proposed under the frame of compressed sensing (CS. To begin with, to prevent the periodical impulses loss and effectively separate periodical impulses from the external noise and additive interference components, the TQWT-MCA method is introduced to divide the raw vibration signal into low-resonance component (LRC, i.e., periodical impulses and high-resonance component (HRC, thus, the periodical impulses are preserved effectively. Then, according to the amplitude range of generated LRC, the step-impulse dictionary atom is designed to match the physical structure of periodical impulses. Furthermore, the periodical impulses and HRC are reconstructed by the sparse Bayesian iteration combined with step-impulse dictionary, respectively, finally, the final reconstructed raw signals are obtained by adding the LRC and HRC, meanwhile, the fidelity of the final reconstructed signals is tested by the envelop spectrum and error analysis, respectively. In this work, the proposed algorithm is applied to simulated signal and engineering multichannel signals of a gearbox with multiple faults. Experimental results demonstrate that the proposed approach significantly improves the reconstructive accuracy compared with the state-of-the-art methods such as non-convex Lq (q = 0.5 regularization, spatiotemporal sparse Bayesian learning (SSBL and L1-norm, etc. Additionally, the processing time, i.e., speed of storage and transmission has increased dramatically, more importantly, the fault characteristics of the gearbox with multiple faults are detected and saved, i.e., the

  7. ITER management advisory committee meeting

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    2001-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held on 23 February in Garching, Germany. The main topics were: the consideration of the report by the Director on the ITER EDA Status, the review of the Work Programme, the review of the Joint Fund, the review of a schedule of ITER meetings, and the arrangements for termination and wind-up of the EDA

  8. Pre-qualification of brazed plasma facing components of divertor target elements for ITER like tokamak application

    International Nuclear Information System (INIS)

    Singh, K.P.; Pandya, Santosh P.; Khirwadkar, S.S.; Patel, Alpesh; Patil, Y.; Buch, J.J.U.; Khan, M.S.; Tripathi, Sudhir; Pandya, Shwetang; Govindrajan, J.; Jaman, P.M.; Rathore, Devendra; Rangaraj, L.; Divakar, C.

    2011-01-01

    Qualification of tungsten (W) and graphite (C) based brazed plasma facing components (PFCs) is an important R and D area in fusion research. Pre-qualification tests for brazed joints between W-CuCrZr and C-CuCrZr using NDT (IR thermography and ultrasonic test) and thermal fatigue test are attempted. Mockups having good quality brazed joints of W and C based PFCs were identified using NDT. Subsequently, thermal fatigue test was performed on the identified mockups. All brazed tiles of W based PFC mockups could withstand thermal fatigue test, however, few tiles of C based PFC mockup were found detached. Thermal analyses of mockups are performed using finite element analysis (ANSYS) software to simulate the thermal hydraulic condition with 10 MW/m 2 uniform heat flux. Details about experimental and computational work are presented here.

  9. Pre-qualification of brazed plasma facing components of divertor target elements for ITER like tokamak application

    Energy Technology Data Exchange (ETDEWEB)

    Singh, K.P., E-mail: kpsingh@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat (India); Pandya, Santosh P.; Khirwadkar, S.S.; Patel, Alpesh; Patil, Y.; Buch, J.J.U.; Khan, M.S.; Tripathi, Sudhir; Pandya, Shwetang; Govindrajan, J. [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat (India); Jaman, P.M.; Rathore, Devendra; Rangaraj, L.; Divakar, C. [Materials Science Division, National Aerospace Laboratories, CSIR, Bangalore, Karnataka (India)

    2011-10-15

    Qualification of tungsten (W) and graphite (C) based brazed plasma facing components (PFCs) is an important R and D area in fusion research. Pre-qualification tests for brazed joints between W-CuCrZr and C-CuCrZr using NDT (IR thermography and ultrasonic test) and thermal fatigue test are attempted. Mockups having good quality brazed joints of W and C based PFCs were identified using NDT. Subsequently, thermal fatigue test was performed on the identified mockups. All brazed tiles of W based PFC mockups could withstand thermal fatigue test, however, few tiles of C based PFC mockup were found detached. Thermal analyses of mockups are performed using finite element analysis (ANSYS) software to simulate the thermal hydraulic condition with 10 MW/m{sup 2} uniform heat flux. Details about experimental and computational work are presented here.

  10. Mechanical thermal and electric measurements on materials and components of the main coils of the Milan superconducting cyclotron

    International Nuclear Information System (INIS)

    Acerbi, E.; Rossi, L.

    1988-01-01

    The coils of the Milan Superconducting Cyclotron are the largest superconducting devices built up to now in Italy and constitute the first superconducting magnet for accelerator in Europe. Because of the large stored energy (more than 40 MJ), of the high stresses and of of the need of reliability, a lot of measurements were carried out as well on materials used for the coils, both on superconducting cable and structural materials, as on the main components of the coils and on two double pancakes prototypes (wound with full copper cable). In this paper the results on these measurements are reported and the results of tests on the prototypes are discussed. The aim is to provide an easy source of data for superconducting coils useful to verify calculations or to improve the performances

  11. Feasibility Study of Cryogenic Cutting Technology by Using a Computer Simulation and Manufacture of Main Components for Cryogenic Cutting System

    International Nuclear Information System (INIS)

    Kim, Sung Kyun; Lee, Dong Gyu; Lee, Kune Woo; Song, Oh Seop

    2009-01-01

    Cryogenic cutting technology is one of the most suitable technologies for dismantling nuclear facilities due to the fact that a secondary waste is not generated during the cutting process. In this paper, the feasibility of cryogenic cutting technology was investigated by using a computer simulation. In the computer simulation, a hybrid method combined with the SPH (smoothed particle hydrodynamics) method and the FE (finite element) method was used. And also, a penetration depth equation, for the design of the cryogenic cutting system, was used and the design variables and operation conditions to cut a 10 mm thickness for steel were determined. Finally, the main components of the cryogenic cutting system were manufactures on the basis of the obtained design variables and operation conditions.

  12. Comparative analysis of the main bioactive components of San-ao decoction and its series of formulations.

    Science.gov (United States)

    Shu, Xiaoyun; Tang, Yuping; Jiang, Chenxue; Shang, Erxing; Fan, Xinshen; Ding, Anwei

    2012-11-01

    A high performance liquid chromatographic (HPLC) method with diode array detection (DAD) was established for simultaneous determination of seven main bioactive components in San-ao decoction and its series of formulae (San-ao decoction, Wu-ao decoction, Qi-ao decoction and Jia-wei San-ao decoction). Seven compounds were analyzed simultaneously with a XTerra C(18) column (4.6 mm × 250 mm, 5 µm) using a linear gradient elution of a mobile phase containing acetonitrile (A) and a buffer solution (0.02 mol/L potassium dihydrogen phosphate and adjusted to pH 3 using phosphoric acid) (B); the flow rate was 1.0 mL/min. The sample was detected with DAD at 210, 254 and 360 nm and the column was maintained at 30 °C. All the compounds showed good linearity (r2 > 0.9984) in the tested concentration range. The precisions were evaluated by intra-day and inter-day tests, and relative standard deviation (R.S.D.) values within the range of 0.83%–2.53% and 0.64%–2.77% were reported, respectively. The recoveries of the quantified compounds were observed to cover a range from 95.34% and 104.82% with R.S.D. values less than 2.72%. The validated method was successfully applied for the simultaneous determination of seven main bioactive components including ephedrine (1), amygdalin (2), liquiritin (3), benzoic acid (4), isoliquiritin (5), formononetin (6) and glycyrrhizic acid (7) in San-ao decoction and its series of formulae. The results also showed a wide variation in the content of the identified active compounds in these samples, which could also be helpful to illustrate the drug interactions after some herbs combined in different formulations.

  13. The ITER divertor concept

    International Nuclear Information System (INIS)

    Janeschitz, G.; Borrass, K.; Federici, G.; Igitkhanov, Y.; Kukushkin, A.; Pacher, H.D.; Pacher, G.W.; Sugihara, M.

    1995-01-01

    The ITER divertor must exhaust most of the alpha particle power and the He ash at acceptable erosion rates. The high recycling regime of the ITER-CDA for present parameters would yield high power loads and erosion rates on conventional targets. Improvement by radiation in the SOL at constant pressure is limited in principle. To permit a higher radiation fraction, the plasma pressure along the field must be reduced by more than a factor 10, reducing also the target ion flux. This pressure reduction can be obtained by strong plasma-neutral interaction below the X-point. Under these conditions T e in the divertor can be reduced to <5 eV along a flame like ionisation front by impurity radiation and CX losses. Downstream of the front, neutrals undergo more CX or i-n collisions than ionisation events, resulting in significant momentum loss via neutrals to the divertor chamber wall. The pressure reduction by this mechanism depends on the along-field length for neutral-plasma interaction, the parallel power flux, the neutral density, the ratio of neutral-neutral collision length to the plasma-wall distance and on the Mach number of ions and neutrals. A supersonic transition in the main plasma-neutral interaction region, expected to occur near the ionisation front, would be beneficial for momentum removal. The momentum transfer fraction to the side walls is calculated: low Knudsen number is beneficial. The impact of the different physics effects on the chosen geometry and on the ITER divertor design and the lifetime of the various divertor components are discussed. ((orig.))

  14. Design improvements and R and D achievements for VV and in-vessel components towards ITER construction

    International Nuclear Information System (INIS)

    Ioki, K.; Barabaschi, P.; Barabash, V.

    2003-01-01

    During the preparation of the procurement specifications for long lead-time items, several detailed vacuum vessel (VV) design improvements are being pursued, such as elimination of the inboard triangular support, adding a separate interspace between inner and outer shells for independent leak detection of field joints, and revising the VV support system to gain a more comfortable structural performance margin. Improvements to the blanket design are also under investigation, an inter-modular key instead of two prismatic keys and a co-axial inlet outlet cooling connection instead of two parallel pipes. One of the most important achievements in the VV R and D has been demonstration of the necessary assembly tolerances. Further development of cutting, welding and nondestructive tests (NDT) for the VV has been continued, and thermal and hydraulic tests have been performed to simulate the VV cooling conditions. In FW/blanket and divertor, full-scale prototypical mock-ups of the FW panel, the blanket shield block, and the divertor components, have been successfully fabricated. These results make us confident in the validity of our design and give us possibilities of alternate fabrication methods. (author)

  15. Design improvements and R and D achievements for VV and In-vessel components towards ITER construction

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Barabaschi, P.; Barabash, V.

    2003-01-01

    There have been several detailed vacuum vessel (VV) design improvements, such as elimination of the inboard triangular support, separate interspace between inner and outer shells for independent leak detection of field joints and revised VV support system to gain a more comfortable margin in the structural performance. The blanket design has been updated; an inter-modular key instead of two prismatic keys and a co-axial inlet-outlet cooling connection instead of two parallel pipes. One of the most important achievements in the VV R and D has been demonstration of the necessary assembly tolerances. Further development of cutting, welding and non destructive tests (NDT) for the VV has been continued, and thermal and hydraulic tests have been performed to simulate the VV cooling conditions. With regard to the R and D for the FW/blanket and divertor, full-scale prototypical mock-ups of the FW panel, the blanket shield block and the divertor components have been successfully fabricated. These results make us confident in the validity of our design and give us possibilities of alternate fabrication methods. (author)

  16. 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

  17. Axial strain localization of CuCrZr tubes during manufacturing of ITER-like mono-block W/Cu components using HIP

    International Nuclear Information System (INIS)

    Zhao, S.X.; Peng, L.J.; Li, Q.; Wang, W.J.; Wei, R.; Qin, S.G.; Shi, Y.L.; Chang, S.P.; Xu, Y.; Liu, G.H.; Wang, T.J.; Luo, G.-N.

    2014-01-01

    Highlights: • Axial cracking and denting of CuCrZr tubes were observed. • Annealing the as-received tubes can alleviate cracking. • Denting results in the formation bonding flaws at the Cu/CuCrZr interfaces. - Abstract: Two forms of axial strain localization of CuCrZr tubes, i.e., cracking and denting, were observed during the manufacturing of ITER-like mono-block W/Cu components for EAST employing hot isostatic pressing (HIP). Microscopic investigations indicate that the occurrence of axial strain localization correlates to the heavily deformed Cu grains and elongated Cr-rich precipitates as well as highly anisotropic microstructures, which impair the circumferential ductility. Annealing the as-received tubes at 600 °C alleviates cracking due to partial recrystallization of Cu grains. However, the annealed tubes are still sensitive to wall thinning (caused by non-uniform polishing or tube bending), which results in denting. Denting may cause bonding flaws at CuCrZr/Cu interfaces and the underlying mechanisms are discussed. To some extent, denting seems do not affect the high heat flux performance of components. In this paper, we demonstrate that testing only the axial mechanical properties is not enough for manufacturers who use HIP or hot radial pressing technologies, especially for those anisotropic tubes

  18. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost

    2012-01-01

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

  19. Power converters for ITER

    CERN Document Server

    Benfatto, I

    2006-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a thermonuclear fusion experiment designed to provide long deuterium– tritium burning plasma operation. After a short description of ITER objectives, the main design parameters and the construction schedule, the paper describes the electrical characteristics of the French 400 kV grid at Cadarache: the European site proposed for ITER. Moreover, the paper describes the main requirements and features of the power converters designed for the ITER coil and additional heating power supplies, characterized by a total installed power of about 1.8 GVA, modular design with basic units up to 90 MVA continuous duty, dc currents up to 68 kA, and voltages from 1 kV to 1 MV dc.

  20. Protective effect of thymoquinone, the main component of Nigella Sativa, against diazinon cardio-toxicity in rats.

    Science.gov (United States)

    Danaei, Gholam Hassan; Memar, Bahram; Ataee, Ramin; Karami, Mohammad

    2018-04-12

    Several studies have shown that oxidative stress and cell damage can occur at very early stages of diazinon (DZN) exposure. The present study was designed to determine the beneficial effect of thymoquinone (Thy), the main component of Nigella sativa (black seed or black cumin), against DZN cardio-toxicity in rats. In the present experimental study, 48 male Wistar rats were randomly divided into six groups: control (corn oil gavages), DZN gavages (20 mg/kg/day), Thy gavages (10 mg/kg/day) and Thy + DVN gavages (2.5, 5 and 10 mg/kg/day). Treatments were continued for 28 days, then the animals were anesthetized by ether and superoxide dismutase (SOD), catalase (CAT), glutathione S-transferase (GST), lactate dehydrogenize (LDH) and glutathione peroxide (GPX) activity was evaluated. In addition, glutathione (GSH) and malondialdehyde (MDA) the heart tissue and creatinephosphokinase-MB (CPK-MB) and troponin (TPI) levels and cholinesterase activity in the blood were evaluated. DZN-induced oxidative damage and elevated the levels of the cardiac markers CK-MB, TPI, MDA and LDH and decreased SOD, CAT and cholinesterase activity and GSH level compared with the control group. Treatment with Thy reduced DZN cardio-toxicity and cholinesterase activity. The success of Thy supplementation against DZN toxicity can be attributed to the antioxidant effects of its constituents. Administration of Thy as a natural antioxidant decreased DZN cardio-toxicity and improved cholinesterase activity in rats through the mechanism of free radical scavenging.

  1. Insecticidal Activity of Essential Oil of Carum Carvi Fruits from China and Its Main Components against Two Grain Storage Insects

    Directory of Open Access Journals (Sweden)

    Shu Shan Du

    2010-12-01

    Full Text Available During our screening program for agrochemicals from Chinese medicinal herbs and wild plants, the essential oil of Carum carvi fruits was found to possess strong contact toxicity against Sitophilus zeamais and Tribolium castaneum adults, with LD50 values of 3.07 and 3.29 mg/adult, respectively, and also showed strong fumigant toxicity against the two grain storage insects with LC50 values of 3.37 and 2.53 mg/L, respectively. The essential oil obtained by hydrodistillation was investigated by GC and GC-MS. The main components of the essential oil were identified to be (R-carvone (37.98% and D-limonene (26.55% followed by α-pinene (5.21, cis-carveol (5.01% and b-myrcene (4.67%. (R-Carvone and D-limonene were separated and purified by silica gel column chromatography and preparative thin layer chromatography, and further identified by means of physicochemical and spectrometric analysis. (R-Carvone and D-limonene showed strong contact toxicity against S. zeamais (LD50 = 2.79 and 29.86 mg/adult and T. castaneum (LD50 = 2.64 and 20.14 mg/adult. (R-Carvone and D-limonene also possessed strong fumigant toxicity against S. zeamais (LC50 = 2.76 and 48.18 mg/L and T. castaneum adults (LC50 = 1.96 and 19.10 mg/L.

  2. Quantitative Proteomic Analysis of Venoms from Russian Vipers of Pelias Group: Phospholipases A₂ are the Main Venom Components.

    Science.gov (United States)

    Kovalchuk, Sergey I; Ziganshin, Rustam H; Starkov, Vladislav G; Tsetlin, Victor I; Utkin, Yuri N

    2016-04-12

    Venoms of most Russian viper species are poorly characterized. Here, by quantitative chromato-mass-spectrometry, we analyzed protein and peptide compositions of venoms from four Vipera species (V. kaznakovi, V. renardi, V. orlovi and V. nikolskii) inhabiting different regions of Russia. In all these species, the main components were phospholipases A₂, their content ranging from 24% in V. orlovi to 65% in V. nikolskii. Altogether, enzyme content in venom of V. nikolskii reached ~85%. Among the non-enzymatic proteins, the most abundant were disintegrins (14%) in the V. renardi venom, C-type lectin like (12.5%) in V. kaznakovi, cysteine-rich venom proteins (12%) in V. orlovi and venom endothelial growth factors (8%) in V. nikolskii. In total, 210 proteins and 512 endogenous peptides were identified in the four viper venoms. They represented 14 snake venom protein families, most of which were found in the venoms of Vipera snakes previously. However, phospholipase B and nucleotide degrading enzymes were reported here for the first time. Compositions of V. kaznakovi and V. orlovi venoms were described for the first time and showed the greatest similarity among the four venoms studied, which probably reflected close relationship between these species within the "kaznakovi" complex.

  3. News from ITER controls - a status report

    International Nuclear Information System (INIS)

    Wallander, A.; Abadie, L.; Di Maio, F.; Evrard, B.; Fourneron, J.M.; Gulati, H.; Hansalia, C.; Journeaux, J.Y.; Kim, C.; Klotz, W.D.; Mahajan, K.; Makijarvi, P; Matsumoto, Y.; Pande, S.; Simrock, S.; Stepanov, D.; Utzel, N.; Vergara, A.; Winter, A.; Yonekawa, I.

    2012-01-01

    Construction of ITER has started at the Cadarache site in southern France. The first buildings are taking shape and more than 60 % of the in-kind procurement has been committed by the seven ITER member states (China, Europe, India, Japan, Korea, Russia and United States). The design and manufacturing of the main components of the machine is now underway all over the world. Each of these components comes with a local control system, which must be integrated in the central control system. The control group at ITER has developed two products to facilitate it; the plant control design handbook (PCDH) and the control, data access and communication (CODAC) core system. PCDH is a document which prescribes the technologies and methods to be used in developing local control systems and sets the rules applicable to the in-kind procurements. CODAC core system is a software package, distributed to all in-kind procurement developers, which implements the PCDH and facilitates the compliance of the local control system. In parallel, the ITER control group is proceeding with the design of the central control system to allow fully integrated and automated operation of ITER. In this paper we report on the progress of the design and technology choices and we discuss justifications of those choices. We also report on the results of some pilot projects aimed at validating the design and technologies. (authors)

  4. Processing of W-Cu functionally graded materials (FGM) through the powder metallurgy route: application as plasma facing components for ITER-like thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Raharijaona, J.J.

    2009-11-01

    The aim of this study was to study and optimize the sintering of W-Cu graded composition materials, for first wall of ITER-like thermonuclear reactor application. The graded composition in the material generates graded functional properties (Functionally Graded Materials - FGM). Rough thermomechanical calculations have shown the interest of W-Cu FGM to improve the lifetime of Plasma Facing Components (PFC). To process W-Cu FGM, powder metallurgy route was analyzed and optimized from W-CuO powder mixtures. The influence of oxide reduction on the sintering of powder mixtures was highlighted. An optimal heating treatment under He/H 2 atmosphere was determined. The sintering mechanisms were deduced from the analysis of the effect of the Cu-content. Sintering of W-Cu materials with a graded composition and grain size has revealed two liquid migration steps: i) capillary migration, after the Cu-melting and, ii) expulsion of liquid, at the end of sintering, from the dense part to the porous part, due to the continuation of W-skeleton sintering. These two steps were confirmed by a model based on capillary pressure calculation. In addition, thermal conductivity measurements were conducted on sintered parts and showed values which gradually increase with the Cu-content. Hardness tests on a polished cross-section in the bulk are consistent with the composition profiles obtained and the differential grain size. (author)

  5. 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

  6. Determining the residual lifetime for the main RBMK components in frame of life extension of power units

    International Nuclear Information System (INIS)

    Baldin, V.D.; Petrov, A.A.; Potapov, A.A.

    2005-01-01

    At present time 11 power units of NPPs with RBMK reactors produce annually 140 billion kWh of electricity, which is about half of all electricity generated at Russian NPPs. Commissioned during the period of 1973 to 1990, they have the specified service life of 30 years. The power units have been regularly upgraded during operation to raise their safety and reliability. As the result now they meet modem requirements in terms of safety and their lifetime can considerably exceed their specified service life. The task of the plant life extension becomes more urgent. The main nonreplaceable RBMK reactor components that determine the reactor life, are graphite stack and metal support structures. The aging of these structures is monitored and the structures residual lifetime is predicted in accordance with procedures approved by regulatory authorities. The lead unit among all RBMK power units is Leningrad NPP Unit 1 commissioned in 1973. A complex of activities was performed for this power unit in 2002-2003 to prove the feasibility of its life extension. In terms of defining the residual lifetime of the reactor components, the activities were performed in three major areas: - examination of the state of the reactor structures; - materials studies of irradiated metal and graphite specimens in hot chambers; - performing the calculation and prediction evaluations. The results of examining the metal structures and studying specimens have shown that: 1)The metal and the welds do not have visible defects. The state of protective coatings is largely satisfactory. 2)The ultimate strength and hardness of steel have grown. 3)The brittle fracture transition temperature shift is at the allowable level. 4)The leak-tightness of the metal structures has not been violated. The results of the graphite stack examination have shown that the graphite is at the hardening stage. No tendency for deterioration of mechanical properties has been recorded. According to the results of measuring the

  7. Effects assessment of 10 functioning years on the main components of the molten salt PCS experimental facility of ENEA

    Science.gov (United States)

    Gaggioli, Walter; Di Ascenzi, Primo; Rinaldi, Luca; Tarquini, Pietro; Fabrizi, Fabrizio

    2016-05-01

    In the frame of the Solar Thermodynamic Laboratory, ENEA has improved CSP Parabolic Trough technologies by adopting new advanced solutions for linear tube receivers and by implementing a binary mixture of molten salt (60% NaNO3 and 40% KNO3) [1] as both heat transfer fluid and heat storage medium in solar field and in storage tanks, thus allowing the solar plants to operate at high temperatures up to 550°C. Further improvements have regarded parabolic mirror collectors, piping and process instrumentation. All the innovative components developed by ENEA, together with other standard parts of the plant, have been tested and qualified under actual solar operating conditions on the PCS experimental facility at the ENEA Casaccia Research Center in Rome (Italy). The PCS (Prova Collettori Solari, i.e. Test of Solar Collectors) facility is the main testing loop built by ENEA and it is unique in the world for what concerns the high operating temperature and the fluid used (mixture of molten salt). It consists in one line of parabolic trough collectors (test section of 100 m long life-size solar collectors) using, as heat transfer fluid, the aforesaid binary mixture of molten salt up to 10 bar, at high temperature in the range 270° and 550°C and a flow rate up to 6.5 kg/s. It has been working since early 2004 [2] till now; it consists in a unique closed loop, and it is totally instrumented. In this paper the effects of over ten years qualification tests on the pressurized tank will be presented, together with the characterization of the thermal losses of the piping of the molten salt circuit, and some observations performed on the PCS facility during its first ten years of operation.

  8. Interaction of calcium silicate hydrates (C-S-H), the main components of cement, with alkaline chlorides, analogy with clays

    International Nuclear Information System (INIS)

    Viallis-Terrisse, H.

    2000-01-01

    This work, belonging to a more general study on the structure and reactivity of cement, deals with the experimental and theoretical analysis of the interaction of alkaline chlorides with calcium silicate hydrates (C-S-H), the main components of cement paste. The interaction of alkaline cations with C-S-H is interfacial, involving both electrostatic and surface complexation mechanisms. The C-S-H surface is constituted of silanol sites, partially dissociated due to the high pH of the interstitial solution. The calcium ions, present in large amounts in the equilibrium solution of C-S-H, constitute potential determining ions for the C-S-H surface. The alkaline ions seem to compete with calcium for the same surface sites. The adsorption isotherms show that caesium presents a better affinity than sodium and lithium for the C-S-H surface. Moreover, solid-state NMR suggests that caesium forms with the surface sites inner-sphere complexes, whereas sodium seems to keep its hydration sphere. These results are in agreement with zeta potential measurements, which let suppose a specific adsorption of caesium ions, and an indifferent behaviour of both other alkaline ions. A model for the C-S-H surface was proposed, from the electric double layer model, and mass action laws expressing the complexation of the different ionic species with the silanol sites. The whole study relies on a structural analogy with smectites, some clays presenting well-known cationic adsorption properties. The structural similarity between both minerals is enhanced by some similarities of reactivity, though significant behaviour differences could also be noted. (author)

  9. Main components of PM10 in an area influenced by a cement plant in Catalonia, Spain: Seasonal and daily variations.

    Science.gov (United States)

    Rovira, Joaquim; Sierra, Jordi; Nadal, Martí; Schuhmacher, Marta; Domingo, José L

    2018-05-01

    Particulate matter (PM) composition has a key role in a wide range of health outcomes, such as asthma, chronic obstructive pulmonary disease, lung cancer, cardiovascular disease, and death, among others. Montcada i Reixac, a municipality located in the Barcelona metropolitan area (Catalonia, Spain), for its location and orography, is an interesting case- study to investigate air pollution. The area is also characterized by the presence of different industrial emission sources, including a cement factory and a large waste management plant, as well as an intense traffic. In this study, PM 10 levels, trace elements, ions, and carbonaceous particles were determined for a long time period (2013-2016) in this highly polluted area. PM 10 samples were collected during six consecutive days in two campaigns (cold and warm) per year. A number of elements (As, Ba, Be, Bi, Ca, Cd, Ce, Co, Cr, Cs, Cu, Dy, Er, Eu, Fe, Ga, Gd, Ge, Hf, Hg, Ho, K, La, Li, Hg, Mg, Mn, Mo, Nb, Nd, Ni, Pb, Pr, Rb, Sb, Sc, Se, Sm, Sn, Sr, Tb, Th, Ti, Tl, U, V, W, Y, Yb, and Zr), ions (Cl - , SO 4 2- , NO 3 - , and NH 4 + ), and carbonaceous content (total carbon, organic plus elemental carbon, and CO 3 2- ), were analysed. These data were used to identify the PM 10 main components: mineral matter, sea spray, secondary inorganic aerosols, organic matter plus elemental carbon, trace elements or indeterminate fraction. Although a clear seasonality (cold vs. warm periods) was found, there were no differences between working days and weekends. Obviously, the cement plant influences the surrounding environment. However, no differences in trace elements related with the cement plant activity (Al, Ca, Ni and V) between weekdays and weekends were noted. However, some traffic-related elements (i.e., Co, Cr, Mn, and Sb) showed significantly higher concentrations in weekdays. Copyright © 2018 Elsevier Inc. All rights reserved.

  10. 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.

  11. The application of iterative transformation factor analysis to resolve multi-component EXAFS spectra of uranium(6) complexes with acetic acid as a function of pH

    International Nuclear Information System (INIS)

    Robberg, A.; Reich, T.

    2002-01-01

    Synchrotron-based EXAFS spectroscopy is a powerful technique to obtain structural information on radionuclide complexes in solution. Depending on the chemical conditions of the samples several radionuclide species can coexist in the solution as is often the case for environmentally related samples. All radionuclide species, which may have different near-neighbour environments, contribute to the measured EXAFS signal. In order to isolate the EXAFS spectra of the individual species (pure spectral components), it is necessary, in a first step, to measure a series of samples where their composition is changed by variation of one physico-chemical parameter (e.g. pH, concentration, etc.). For the spectral decomposition it is necessary that the EXAFS signal change as a function of the chosen physico-chemical parameter. In a second step, the series of EXAFS spectra is analysed with Eigen analysis and Iterative Transformation Factor Analysis (ITFA). As a result of the ITFA one obtains: a) for each sample the relative concentration of the structural distinguishable species and b) their corresponding pure spectral components. From the information obtained in a), one can construct a speciation diagram. The pure spectral components contain the structural information of the individual species, which can be extracted by conventional EXAFS analysis. To evaluate our ITFA algorithm for EXAFS analysis of mixtures, we prepared a series of eight solution samples of 0.05 M uranium(VI) and 1 M acetate (Ac) in the pH range of 0.1 to 4.5. From thermodynamic constants it is known that under these conditions up to four species can occur: uranyl hydrate, and the 1:1, 1:2 and 1:3 complexes of uranyl acetate. The uranium L III -edge EXAFS spectra were measured at room temperature in transmission mode at the Rossendorf Beamline (ROBL) at the ESRF. The average bond length between uranium and the equatorial oxygen atoms (O eq ) increases from 2.40 to 2.46 angstrom with increasing pH. This increase

  12. ITER plasma facing materials. Some critical considerations

    International Nuclear Information System (INIS)

    Barabash, V.; Dietz, K.J.; Federici, G.; Janeschitz, G.; Matera, R.; Tanaka, S.

    1995-01-01

    The description of current status with the choice of materials for ITER plasma facing components is presented. The main problem with lifetime of divertor elements is the particle and energy-induced erosion of armour materials. A solution for the first operation phase consists in using Be as an armour for the first wall and the divertor, however other possible materials (e.g. W) could be considered. (orig.)

  13. Progress and achievements of the ITER L-4 blanket project

    International Nuclear Information System (INIS)

    Daenner, W.; Toschi, R.; Cardella, A.

    1999-01-01

    The L-4 Blanket Project embraces the R and D of the ITER Shielding Blanket, and its main objective is the fabrication of prototype components. This paper summarises the main conclusions from the materials R and D and the development of technologies which were required for the prototype specifications and manufacturing. The main results of the ongoing testing activities, and of the component manufacture are outlined.The main objectives of the project have been achieved including improvements of the material properties and of joining technologies, which resulted in good component quality and high performance in qualification tests. (author)

  14. Progress and achievements of the ITER L-4 blanket project

    International Nuclear Information System (INIS)

    Daenner, W.; Toschi, R.; Cardella, A.

    2001-01-01

    The L-4 Blanket Project embraces the R and D of the ITER Shielding Blanket, and its main objective is the fabrication of prototype components. This paper summarises the main conclusions from the materials R and D and the development of technologies which were required for the prototype specifications and manufacturing. The main results of the ongoing testing activities, and of the component manufacture are outlined. The main objectives of the project have been achieved including improvements of the material properties and of joining technologies, which resulted in good component quality and high performance in qualification tests. (author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    2001-01-01

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

  17. [Determination of six main components in compound theophylline tablet by convolution curve method after prior separation by column partition chromatography

    Science.gov (United States)

    Zhang, S. Y.; Wang, G. F.; Wu, Y. T.; Baldwin, K. M. (Principal Investigator)

    1993-01-01

    On a partition chromatographic column in which the support is Kieselguhr and the stationary phase is sulfuric acid solution (2 mol/L), three components of compound theophylline tablet were simultaneously eluted by chloroform and three other components were simultaneously eluted by ammonia-saturated chloroform. The two mixtures were determined by computer-aided convolution curve method separately. The corresponding average recovery and relative standard deviation of the six components were as follows: 101.6, 1.46% for caffeine; 99.7, 0.10% for phenacetin; 100.9, 1.31% for phenobarbitone; 100.2, 0.81% for theophylline; 99.9, 0.81% for theobromine and 100.8, 0.48% for aminopyrine.

  18. ITER licensing

    International Nuclear Information System (INIS)

    Gordon, C.W.

    2005-01-01

    ITER was fortunate to have four countries interested in ITER siting to the point where licensing discussions were initiated. This experience uncovered the challenges of licensing a first of a kind, fusion machine under different licensing regimes and helped prepare the way for the site specific licensing process. These initial steps in licensing ITER have allowed for refining the safety case and provide confidence that the design and safety approach will be licensable. With site-specific licensing underway, the necessary regulatory submissions have been defined and are well on the way to being completed. Of course, there is still work to be done and details to be sorted out. However, the informal international discussions to bring both the proponent and regulatory authority up to a common level of understanding have laid the foundation for a licensing process that should proceed smoothly. This paper provides observations from the perspective of the International Team. (author)

  19. Extending ITER materials design to welded joints

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.-A.F. [DMN/Dir, CEA/Saclay, Commissariat a l' Energie Atomique, 91191 Gif sur Yvette cedex (France)]. E-mail: tavassoli@cea.fr

    2007-08-01

    This paper extends the ITER materials properties documentation to weld metals and incorporates the needs of Test Blanket Modules for higher temperature materials properties. Since the main structural material selected for ITER is type 316L(N)-IG, the paper is focused on weld metals and joining techniques for this steel. Materials properties data are analysed according to the French design and construction rules for nuclear components (RCC-MR) and design allowables are equally derived using the same rules. Particular attention is paid to the type of weld metal, to the type and position of welding and their influence on the materials properties data and design allowables. The primary goal of this work, starting with 19-12-2 weld metal, is to produce comprehensive materials properties documentations that when combined with codification and inspection documents would satisfy ITER licensing needs. As a result, structural stability and capability of welded joints during manufacturing of ITER components and their subsequent service, including the effects of irradiation and eventual incidental or accidental situations, are also covered.

  20. Nuclear analysis for ITER

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  1. ITER management advisory committee meeting

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    2001-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held in Vienna on 16 July 2001. It was the last MAC Meeting and the main topics were consideration of the report by the Director on the ITER EDA status, review of the Work Programme, review of the Joint Fund and arrangements for termination and wind-up of the EDA

  2. The ITER activity

    International Nuclear Information System (INIS)

    Glass, A.J.

    1991-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is a collaboration among four parties, the United States, the Soviet Union, Japan, and the European Communities, to demonstrate the scientific and technological feasibility of fusion power for peaceful purposes. ITER will demonstrate this through the construction of a tokamak fusion reactor capable of generating 1000 megawatts of fusion power. The ITER project has three missions, as follows: (1) Physics mission -- to demonstrate ignition and controlled burn, with pulse durations from 200 to 1000 S; (2) Technology mission -- to demonstrate the technologies essential to a reactor in an integrated system, operating with high reliability and availability in pulsed operation, with steady-state operation as the ultimate goal; and (3) Testing mission -- to test nuclear and high-heat-flux components at flux levels for 1 mw/m 2 , and fluences of order 1 mw-yr/m 2

  3. How insects overcome two-component plant chemical defence: plant β-glucosidases as the main target for herbivore adaptation.

    Science.gov (United States)

    Pentzold, Stefan; Zagrobelny, Mika; Rook, Fred; Bak, Søren

    2014-08-01

    Insect herbivory is often restricted by glucosylated plant chemical defence compounds that are activated by plant β-glucosidases to release toxic aglucones upon plant tissue damage. Such two-component plant defences are widespread in the plant kingdom and examples of these classes of compounds are alkaloid, benzoxazinoid, cyanogenic and iridoid glucosides as well as glucosinolates and salicinoids. Conversely, many insects have evolved a diversity of counteradaptations to overcome this type of constitutive chemical defence. Here we discuss that such counter-adaptations occur at different time points, before and during feeding as well as during digestion, and at several levels such as the insects’ feeding behaviour, physiology and metabolism. Insect adaptations frequently circumvent or counteract the activity of the plant β-glucosidases, bioactivating enzymes that are a key element in the plant’s two-component chemical defence. These adaptations include host plant choice, non-disruptive feeding guilds and various physiological adaptations as well as metabolic enzymatic strategies of the insect’s digestive system. Furthermore, insect adaptations often act in combination, may exist in both generalists and specialists, and can act on different classes of defence compounds. We discuss how generalist and specialist insects appear to differ in their ability to use these different types of adaptations: in generalists, adaptations are often inducible, whereas in specialists they are often constitutive. Future studies are suggested to investigate in detail how insect adaptations act in combination to overcome plant chemical defences and to allow ecologically relevant conclusions.

  4. Prediction of radionuclide accumulation in main ecosystem components of NPP cooling water reservoirs and assessment of acceptable radionuclide disposal into water reservoir

    International Nuclear Information System (INIS)

    Egorov, Yu.A.; Kazakov, S.V.

    1987-01-01

    The problems of prediction of radionuclide accumulation in ecosystem main components of NPP cooling water-reservoirs (CWR) and assessment of radionuclide acceptable disposal into water reservoir are considered. Two models are nessecary for the calculation technique: model of radionuclide migration and accumulation in CWR ecosystem components and calculation model of population dose commitment due to water consumption (at the public health approach to the normalization of the NPP radioactive effect on CWC) or calculation model of dose commitment on hydrocenosis components (at the ecological approach to the normalization). Analytical calculations and numerical calculation results in the model CWC, located in the USSR middle region, are presented

  5. 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

  6. Technical basis for the ITER detailed design report, cost review and safety analysis (DDR)

    International Nuclear Information System (INIS)

    1997-01-01

    The ITER Detailed Design Report (DDR), Cost Review and Safety Analysis is the 3rd major milestone representing the progress made in the ITER Engineering Design Activities. With the approval of the Interim Design Report (IDR), it has been possible to freeze the main concepts and system approaches for ITER and to develop the design in more detail for the individual components and sub-systems. This report, although designed to be fully understandable as a separate document, focusses particularly on the main changes since the IDR

  7. Technical basis for the ITER detailed design report, cost review and safety analysis (DDR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The ITER Detailed Design Report (DDR), Cost Review and Safety Analysis is the 3rd major milestone representing the progress made in the ITER Engineering Design Activities. With the approval of the Interim Design Report (IDR), it has been possible to freeze the main concepts and system approaches for ITER and to develop the design in more detail for the individual components and sub-systems. This report, although designed to be fully understandable as a separate document, focusses particularly on the main changes since the IDR. Refs, figs, tabs

  8. Comparative analysis of the main bioactive components of Xin-Sheng-Hua granule and its single herbs by ultrahigh performance liquid chromatography with tandem mass spectrometry.

    Science.gov (United States)

    Pang, Hanqing; Wang, Jun; Tang, Yuping; Xu, Huiqin; Wu, Liang; Jin, Yi; Zhu, Zhenhua; Guo, Sheng; Shi, Xuqin; Huang, Shengliang; Sun, Dazheng; Duan, Jin-Ao

    2016-11-01

    Xin-Sheng-Hua granule, a representative formula for postpartum hemorrhage, has been used clinically to treat postpartum diseases. Its main bioactive components comprise aromatic acids, phthalides, alkaloids, flavonoids, and gingerols among others. To investigate the changes in main bioactive constituents in its seven single herbs before and after compatibility, a rapid, simple, and sensitive method was developed for comparative analysis of 27 main bioactive components by using ultrahigh-performance liquid chromatography with triple quadrupole electrospray tandem mass spectrometry for the first time. The sufficient separation of 27 target constituents was achieved on a Thermo Scientific Hypersil GOLD column (100 mm × 3 mm, 1.9 μm) within 20 min under the optimized chromatographic conditions. Compared with the theoretical content, the observed content of each analyte showed remarkable differences in Xin-Sheng-Hua granule except thymine, p-coumaric acid, senkyunolide I, senkyunolide H, and ligustilide; the total contents of 27 components increased significantly, and the content variation degrees for the different components were gingerols > flavonoids > aromatic acids > alkaloids > phthalides. The results could provide a good reference for the quality control of Xin-Sheng-Hua granule and might be helpful to interpret the drug interactions based on variation of bioactive components in formulae. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  9. Principal component analysis of the main factors of line intensity enhancements observed in oscillating direct current plasma

    International Nuclear Information System (INIS)

    Stoiljkovic, Milovan M.; Pasti, Igor A.; Momcilovic, Milos D.; Savovic, Jelena J.; Pavlovic, Mirjana S.

    2010-01-01

    Enhancement of emission line intensities by induced oscillations of direct current (DC) arc plasma with continuous aerosol sample supply was investigated using multivariate statistics. Principal component analysis (PCA) was employed to evaluate enhancements of 34 atomic spectral lines belonging to 33 elements and 35 ionic spectral lines belonging to 23 elements. Correlation and classification of the elements were done not only by a single property such as the first ionization energy, but also by considering other relevant parameters. Special attention was paid to the influence of the oxide bond strength in an attempt to clarify/predict the enhancement effect. Energies of vaporization, atomization, and excitation were also considered in the analysis. In the case of atomic lines, the best correlation between the enhancements and first ionization energies was obtained as a negative correlation, with weak consistency in grouping of elements in score plots. Conversely, in the case of ionic lines, the best correlation of the enhancements with the sum of the first ionization energies and oxide bond energies was obtained as a positive correlation, with four distinctive groups of elements. The role of the gas-phase atom-oxide bond energy in the entire enhancement effect is underlined.

  10. Purified Cannabidiol, the main non-psychotropic component of Cannabis sativa, alone, counteracts neuronal apoptosis in experimental multiple sclerosis.

    Science.gov (United States)

    Giacoppo, S; Soundara Rajan, T; Galuppo, M; Pollastro, F; Grassi, G; Bramanti, P; Mazzon, E

    2015-12-01

    Multiple Sclerosis (MS) is a global concern disease leading to a progressive, chronic and demyelinating condition, affecting the central nervous system (CNS). The pathology has an inflammatory/autoimmune origin; nevertheless, neuronal cell death mechanisms are not to be underestimated. The present study was designed to test the effects of intraperitoneal administration of cannabidiol (CBD), the main non-psychotropic cannabinoid of Cannabis sativa (CS), in an experimental model of MS. The aim is to evaluate the capability of CBD administration to thwart the cascade of mediators involved in MS-induced apoptosis. Experimental Autoimmune Encephalomyelitis (EAE) was induced by immunization with myelin oligodendroglial glycoprotein (MOG)35-55 peptide in mice. After immunization, mice were observed daily for signs of EAE and weight loss. Disease signs were evaluated using a standardized scoring system. Immunohistochemical and Western blot assessments of key apoptotic markers reveal that CBD treatment is able to avoid Fas pathway activation, phospho-ERK p42/44 and cleaved caspase-3 triggering as well as alterations in mitochondrial permeability due to Bax/Bcl-2 unbalance. Moreover, CBD interferes with p53-p21 axis activation. As results, the absence of tissue apobody formation in spinal cord tissues of EAE-mice treated with CBD was established. Most of therapeutic properties of CS are currently ascribed to the psychotropic effects of phenylterpenoid delta-9 tetrahydrocannabinol. We have demonstrated that, alone, purified CBD possesses an anti-apoptotic power against the neurodegenerative processes underlying MS development. This represents an interesting new profile of CBD that could lead to its introduction in the clinical management of MS.

  11. ITER ITA newsletter. No. 2, March 2003

    International Nuclear Information System (INIS)

    2003-05-01

    This ITER ITA newsletter contains concise information about the JASS (Joint Assessment of Specific Sites) final report. The main conclusion is that all four sites are sound and fully capable to respond to all ITER Site Requirements and Design assumptions, as approved by the ITER Council in its January 2000 Meeting

  12. Effects of growing location on the production of main active components and antioxidant activity of Dasiphora fruticosa (L.) Rydb. by chemometric methods.

    Science.gov (United States)

    Liu, Wei; Wang, Dongmei; Hou, Xiaogai; Yang, Yueqin; Xue, Xian; Jia, Qishi; Zhang, Lixia; Zhao, Wei; Yin, Dongxue

    2018-05-17

    Traditional Chinese medicine (TCM) plays a very important role in the health system of China. The content and activity of active component are main indexes that evaluate the quality of TCM, however they may vary with environmental factors in their growing locations. In this study, effects of environmental factors on the contents of active components and antioxidant activity of Dasiphora fruticosa from the five main production areas of China were investigated. The contents of tannin, total flavonoid and rutin were determined and varied within the range of 7.65-10.69%, 2.30-5.39% and 0.18-0.81%, respectively. Antioxidant activity was determined by DPPH assay, with the DPPH IC 50 values ranged from 8.791 to 32.534μg mL -1 . In order to further explore the cause of these significant geographical variations, the chemometric methods including correlation analysis, principal component analysis, gray correlation analysis, and path analysis were conducted. The results showed environmental factors had significant effect on the active component contents and antioxidant activity. Rapidly available phosphorus (RAP) and rapidly available nitrogen (RAN) were common dominant factors, and a significant positive correlation was observed between RAP and active components and antioxidant activity (Pcomponents and strong antioxidant activity, Bange in Tibet and Geermu in Qinghai Province was selected as a favorable growing location, respectively. This article is protected by copyright. All rights reserved. This article is protected by copyright. All rights reserved.

  13. Involvement of the EU industry in ITER EDA

    International Nuclear Information System (INIS)

    Bogusch, E.

    2001-01-01

    Since the fifties, European industry has been involved in research and development in the field of nuclear fusion as a potential future source of energy. Early contributions mainly included deliveries of plant components and services to experimental facilities. In the Engineering Design Activities (EDA) phase of the planned multinational International Thermonuclear Experimental Reactor (ITER) in 1993 to 2001 this commitment of industry was intensified. Industries from seven European countries participated in the project with various contributions, e.g., in the development, design, and manufacture of components, and in the development of methods of planning and executing the complex ITER project. These activities were accompanied by an extensive R and D program. e.g., about materials and methods of manufacturing ITER components. In this way, European industry made an important contribution to the further development of nuclear fusion within the framework of ITER EDA activities, and will be able to continue this work intensively in the expected ITER construction phase to follow. (orig.) [de

  14. ITER In-Cryostat inspection and repair feasibility studies

    International Nuclear Information System (INIS)

    Reich, J.; Cordier, J.-J.; Houtte, D. van; Evrard, D.; Mercier, E.; Popa, T.; Doshi, B.

    2011-01-01

    The ITER In-Cryostat maintenance study is an important precondition to guarantee the operation over the ITER lifetime. The ITER operation is subdivided mainly into two phases: 1.Hydrogen phase (non-nuclear operation phase). 2.Deuterium/Tritium phase (nuclear DT phase). The commissioning phase includes the initial phase of assembly. Within the first phase the ITER components will be tested; afterwards they will go into operation. The In-Cryostat maintenance shall facilitate all operations that could be required by In-Cryostat systems and the Cryostat itself. In cases of failures or unlikely events (e.g. earthquakes) it is necessary to provide man and tool access to In-Cryostat components. Overall functions which have to be implemented are: ·Inspection of components including leak localization (helium, water, air). ·Repair and replacement of component (instrumentation, parts or complete components). ·Regulatory inspections. It is presumed that most of component failure would occur at the beginning of the operational phase. This failure rate is expected to be very unlikely when ITER is being operating during the nuclear phase. For maintenance activities it is assumed that: ·The intervention frequency on each component is limited during its lifetime (e.g. inspections/repair during global shutdown). ·Most of these interventions will be required during the inactive phase. According to ALARA (As Low as Reasonable Achievable) rules maintenance activities will be planned in order to minimize the required human interventions during the active phase. Different tools have to be designed to perform the maintenance actions. As there are quiet all heavy components to be handled and removed, humans cannot perform the work without semi hands-on tools. The required permanent fixtures and tools are considered and pre-designed.

  15. ITER In-Cryostat inspection and repair feasibility studies

    Energy Technology Data Exchange (ETDEWEB)

    Reich, J., E-mail: Jens.Reich@iter.org [ITER Organization, CS 90 046, 13115 St Paul lez Durance Cedex (France); Cordier, J.-J.; Houtte, D. van [ITER Organization, CS 90 046, 13115 St Paul lez Durance Cedex (France); Evrard, D. [Sogeti High Tech, 180 rue Rene Descartes, 13857 Aix en Provence (France); Mercier, E. [AREVA CNIM KAH System Engineering Support, CS 50497, 13593 Aix en Provence Cedex 3 (France); Popa, T.; Doshi, B. [ITER Organization, CS 90 046, 13115 St Paul lez Durance Cedex (France)

    2011-10-15

    The ITER In-Cryostat maintenance study is an important precondition to guarantee the operation over the ITER lifetime. The ITER operation is subdivided mainly into two phases: 1.Hydrogen phase (non-nuclear operation phase). 2.Deuterium/Tritium phase (nuclear DT phase). The commissioning phase includes the initial phase of assembly. Within the first phase the ITER components will be tested; afterwards they will go into operation. The In-Cryostat maintenance shall facilitate all operations that could be required by In-Cryostat systems and the Cryostat itself. In cases of failures or unlikely events (e.g. earthquakes) it is necessary to provide man and tool access to In-Cryostat components. Overall functions which have to be implemented are: {center_dot}Inspection of components including leak localization (helium, water, air). {center_dot}Repair and replacement of component (instrumentation, parts or complete components). {center_dot}Regulatory inspections. It is presumed that most of component failure would occur at the beginning of the operational phase. This failure rate is expected to be very unlikely when ITER is being operating during the nuclear phase. For maintenance activities it is assumed that: {center_dot}The intervention frequency on each component is limited during its lifetime (e.g. inspections/repair during global shutdown). {center_dot}Most of these interventions will be required during the inactive phase. According to ALARA (As Low as Reasonable Achievable) rules maintenance activities will be planned in order to minimize the required human interventions during the active phase. Different tools have to be designed to perform the maintenance actions. As there are quiet all heavy components to be handled and removed, humans cannot perform the work without semi hands-on tools. The required permanent fixtures and tools are considered and pre-designed.

  16. Activation analyses updating the ITER radioactive waste assessment

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  17. Assessment of the flotability of chalcopyrite, molybdenite and pyrite using biosolids and their main components as collectors for greening the froth flotation of copper sulphide ores.

    OpenAIRE

    Sobarzo, Francisco; Herrera Urbina, Ronaldo; Higueras Higueras, Pablo Leon; Sáez Navarrete, César; Godoy Faúndez, Alex; Reyes Bozo, Lorenzo; Vásquez Bestagno, Jorge

    2014-01-01

    Biosolids and representative compounds of their main components ? humic acids, sugars, and proteins ? have been tested as possible environment-friendly collectors and frothers for the flotation of copper sulphide ores. The floatability of chalcopyrite and molybdenite ? both valuable sulphide minerals present in these ores ? as well as non-valuable pyrite was assessed through Hallimond tube flotation tests. Humic acids exhibit similar collector ability for chalcopyrite and molybdenite as that ...

  18. Effect of SPL (Spent Pot Liner) and its main components on root growth, mitotic activity and phosphorylation of Histone H3 in Lactuca sativa L.

    Science.gov (United States)

    Freitas, Aline Silva; Fontes Cunha, Isabela Martinez; Andrade-Vieira, Larissa Fonseca; Techio, Vânia Helena

    2016-02-01

    Spent Pot Liner (SPL) is a solid waste from the aluminum industry frequently disposed of in industrial landfills; it can be leached and contaminate the soil, sources of drinking water and plantations, and thus may pose a risk to human health and to ecosystems. Its composition is high variable, including cyanide, fluoride and aluminum salts, which are highly toxic and environmental pollutants. This study evaluated the effect of SPL and its main components on root growth and the mitosis of Lactuca sativa, by investigating the mechanisms of cellular and chromosomal alterations with the aid of immunolocalization. To this end, newly emerged roots of L. sativa were exposed to SPL and its main components (solutions of cyanide, fluoride and aluminum) and to calcium chloride (control) for 48h. After this, root length was measured and cell cycle was examined by means of conventional cytogenetics and immunolocalization. Root growth was inhibited in the treatments with SPL and aluminum; chromosomal and nuclear alterations were observed in all treatments. The immunolocalization evidenced normal dividing cells with regular temporal and spatial distribution of histone H3 phosphorylation at serine 10 (H3S10ph). However, SPL and its main components inhibited the phosphorylation of histone H3 at serine 10, inactivated pericentromeric regions and affected the cohesion of sister chromatids, thus affecting the arrangement of chromosomes in the metaphase plate and separation of chromatids in anaphase. In addition, these substances induced breaks in pericentromeric regions, characterized as fragile sites. Copyright © 2015 Elsevier Inc. All rights reserved.

  19. 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

  20. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  1. 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.)

  2. The ITER Neutral Beam Test Facility towards SPIDER operation

    Science.gov (United States)

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

    2017-08-01

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

  3. ITER EDA newsletter. V. 3, no. 2

    International Nuclear Information System (INIS)

    1994-02-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the Fifth ITER Council Meeting held in Garching, Germany, 27-28 January 1994, a visit (28 January 1994) of an international group of Harvard Fellows to the San Diego Joint Work Site, the Inauguration Ceremony of the EC-hosted ITER joint work site in Garching (28 January 1994), on an ITER Technical Meeting on Assembly and Maintenance held in Garching, Germany, January 19-26, 1994, and a report on a Technical Committee Meeting on radiation effects on in-vessel components held in Garching, Germany, November 15-19, 1993, as well as an ITER Status Report

  4. 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.)

  5. Divertor development for ITER

    International Nuclear Information System (INIS)

    Janeschitz, G.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Matera, R.; Martin, E.; Parker, R.; Tivey, R.; Pacher, H.D.

    1998-01-01

    The requirements for the ITER divertor design, i.e. power and He ash exhaust, neutral leakage control, lifetime, disruption load resistance and exchange by remote handling, are described in this paper. These requirements and the physics requirements for detached and semi-attached operation result in the vertical target configuration. This is realised by a concept incorporating 60 cassettes carrying the high heat flux components. The armour choice for these components is CFC monoblock in the strike zone near at the lower part of the vertical target, and a W brush elsewhere. Cooling is by swirl tubes or hypervapotrons depending on the component. The status of the heat sink and joining technology R and D is given. Finally, the resulting design of the high heat flux components is presented. (orig.)

  6. ITER council proceedings: 2001

    International Nuclear Information System (INIS)

    2001-01-01

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

  7. 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.)

  8. 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

  9. Development of 20 kW input power coupler for 1.3 GHz ERL main linac. Component test at 30 kW IOT test stand

    International Nuclear Information System (INIS)

    Sakai, Hiroshi; Umemori, Kensei; Sakanaka, Shogo; Takahashi, Takeshi; Furuya, Takaaki; Shinoe, Kenji; Ishii, Atsushi; Nakamura, Norio; Sawamura, Masaru

    2009-01-01

    We started to develop an input coupler for a 1.3 GHz ERL superconducting cavity. Required input power is about 20 kW for the cavity acceleration field of 20 MV/m and the beam current of 100 mA in energy recovery operation. The input coupler is designed based on the STF-BL input coupler and some modifications are applied to the design for the CW 20 kW power operation. We fabricated input coupler components such as ceramic windows and bellows and carried out the high-power test of the components by using a 30 kW IOT power source and a test stand constructed for the highpower test. In this report, we mainly describe the results of the high-power test of ceramic window and bellows. (author)

  10. Process and overview of diagnostics integration in ITER ports

    International Nuclear Information System (INIS)

    Drevon, J.M.; Walsh, M.; Andrew, P.; Barnsley, R.; Bertalot, L.; Bock, M. de; Bora, D.; Bouhamou, R.; Direz, M.F.; Encheva, A.; Fang, T.; Feder, R.; Giacomin, T.; Hellermann, M. von; Jakhar, S.; Johnson, D.; Kaschuk, Y.; Kusama, Y.; Lee, H.G.; Levesy, B.

    2013-01-01

    Highlights: ► An overview of the Port Integration hardware for tenant system hosting inside ITER diagnostics ports is given. ► The main challenges for diagnostic port integration engineering are presented. ► The actions taken for a common modular approach and a coordinated design are detailed. -- Abstract: ITER will have a set of 45 diagnostics to ensure controlled operation. Many of them are integrated in the ITER ports. This paper addresses the integration process of the diagnostic systems and the approach taken to enable coordinated progress. An overview of the Port Integration hardware introduces the various structures needed for hosting tenant systems inside ITER diagnostics ports. The responsibilities of the different parties involved (ITER Organization and the Domestic Agencies) are outlined. The main challenges for diagnostic port integration engineering are summarized. The plan for a common approach to design and manufacture of the supporting structures, in particular the Port Plug is detailed. A coordinated design including common components and a common approach for neutronic analyses is proposed. One particular port, the equatorial port 11, is used to illustrate the approach

  11. ITER technical basis

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-01-01

    Following on from the Final Report of the EDA(DS/21), and the summary of the ITER Final Design report(DS/22), the technical basis gives further details of the design of ITER. It is in two parts. The first, the Plant Design specification, summarises the main constraints on the plant design and operation from the viewpoint of engineering and physics assumptions, compliance with safety regulations, and siting requirements and assumptions. The second, the Plant Description Document, describes the physics performance and engineering characteristics of the plant design, illustrates the potential operational consequences foe the locality of a generic site, gives the construction, commissioning, exploitation and decommissioning schedule, and reports the estimated lifetime costing based on data from the industry of the EDA parties.

  12. ITER technical basis

    International Nuclear Information System (INIS)

    2002-01-01

    Following on from the Final Report of the EDA(DS/21), and the summary of the ITER Final Design report(DS/22), the technical basis gives further details of the design of ITER. It is in two parts. The first, the Plant Design specification, summarises the main constraints on the plant design and operation from the viewpoint of engineering and physics assumptions, compliance with safety regulations, and siting requirements and assumptions. The second, the Plant Description Document, describes the physics performance and engineering characteristics of the plant design, illustrates the potential operational consequences foe the locality of a generic site, gives the construction, commissioning, exploitation and decommissioning schedule, and reports the estimated lifetime costing based on data from the industry of the EDA parties

  13. Rapid determination of main components by means of flame-atomic-absorption spectrometry for chromium, silicon and tungsten in CrSiW materials

    International Nuclear Information System (INIS)

    Mueller, E.; Stahlberg, R.

    1985-01-01

    The application of Flame-Atomic-Absorption Spectrometry (FAAS) for determining chromium, silicon and tungsten in CrSiW materials is described. The FAAS determinations of the main components are shown under optimum conditions. Sufficient precision and reliability have been achieved for routine analysis. The application of a mixture of acids for preparing CrSiW solutions is proposed. The preparation of samples is discussed in detail. Optimum conditions are recommended for determining chromium, silicon and tungsten using one solution only. (orig.) [de

  14. Optimization of Extraction Process for Antidiabetic and Antioxidant Activities of Kursi Wufarikun Ziyabit Using Response Surface Methodology and Quantitative Analysis of Main Components.

    Science.gov (United States)

    Edirs, Salamet; Turak, Ablajan; Numonov, Sodik; Xin, Xuelei; Aisa, Haji Akber

    2017-01-01

    By using extraction yield, total polyphenolic content, antidiabetic activities (PTP-1B and α -glycosidase), and antioxidant activity (ABTS and DPPH) as indicated markers, the extraction conditions of the prescription Kursi Wufarikun Ziyabit (KWZ) were optimized by response surface methodology (RSM). Independent variables were ethanol concentration, extraction temperature, solid-to-solvent ratio, and extraction time. The result of RSM analysis showed that the four variables investigated have a significant effect ( p analysis of effective part of KWZ was characterized via UPLC method, 12 main components were identified by standard compounds, and all of them have shown good regression within the test ranges and the total content of them was 11.18%.

  15. Hydrophilic compounds in culture filtrates of Fusarium oxysporum f. sp. cubense GCV [01210] induce protection to banana leave toward a main pathogen phytotoxic component

    Directory of Open Access Journals (Sweden)

    Nayanci Portal González

    2014-07-01

    Full Text Available Panama disease caused by Fusarium oxysporum f. sp. cubense (Foc, is among the most important diseases in Musa spp. Foc is a necrotrophic fungus, their phytotoxins play a role in disease development. Previously culture filtrate (FCC 15 days incubation with differential phytotoxic activity against two Musa cultivars was obtained. From this, the main fraction with nonspecific phytotoxic activity against both cultivars was purified. In this study, the biological activity of the aqueous phase and the main phytotoxic fraction purified from organic extract of Fusarium oxysporum f. sp. cubense VCG [01210] Race 1 FCC was determined on banana leaves of cv. `Gros Michel' (susceptible and `FHIA-01' (resistant. Foc FCC phytotoxic effect was confirmed. The aqueous phase showed no phytotoxic activity on both cultivars, while the simultaneous application of the aqueous phase with the main phytotoxic fraction induced a differential response of tissues in susceptible and resistant cultivars evaluated. The results indicated that the compounds present in the aqueous phase are required to induce the protection of leaf tissue against phytotoxic main component of the pathogen. Key words: culture filtrate, Panama disease, resistant, susceptible

  16. Preparation of acceptance tests and criteria for the Test Blanket Systems to be operated 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, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Cuquel, B. [AIRBUS Defence and Space S.A.S., 13115 Saint Paul Lez Durance (France); Demange, D.; Ghidersa, B.-E. [Karlsruhe Institute of Technology, Karlsruhe (Germany); Giancarli, L.M.; Iseli, M.; Jourdan, T. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Pascal, R.; Ring, W. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • Initial guideline for acceptance testing and acceptance criteria for Test Blanket Systems in ITER. • These tests complement those required by the applicable codes and standards, and regulations. • Completion of TBS manufacture will be followed by Factory Acceptance Testing, prior to shipment. • Next steps are “Reception Inspection Tests”, and on-site pre-installation and components tests. • This guideline allows the detailing of the TBS specific test plans and their scheduling. - Abstract: This paper describes the main acceptance criteria and required acceptance tests for the components of the six Test Blanket Systems to be installed and operated in ITER. It summarizes the guide-line toward the establishment of detailed test plans for the TBS, starting from the end-product at the ITER Members factories, and to generally define the type of tests that have to be performed on the ITER site after shipment, and/or prior to the systems final commissioning phase.

  17. 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

  18. 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

  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. Searching for the main anti-bacterial components in artificial Calculus bovis using UPLC and microcalorimetry coupled with multi-linear regression analysis.

    Science.gov (United States)

    Zang, Qing-Ce; Wang, Jia-Bo; Kong, Wei-Jun; Jin, Cheng; Ma, Zhi-Jie; Chen, Jing; Gong, Qian-Feng; Xiao, Xiao-He

    2011-12-01

    The fingerprints of artificial Calculus bovis extracts from different solvents were established by ultra-performance liquid chromatography (UPLC) and the anti-bacterial activities of artificial C. bovis extracts on Staphylococcus aureus (S. aureus) growth were studied by microcalorimetry. The UPLC fingerprints were evaluated using hierarchical clustering analysis. Some quantitative parameters obtained from the thermogenic curves of S. aureus growth affected by artificial C. bovis extracts were analyzed using principal component analysis. The spectrum-effect relationships between UPLC fingerprints and anti-bacterial activities were investigated using multi-linear regression analysis. The results showed that peak 1 (taurocholate sodium), peak 3 (unknown compound), peak 4 (cholic acid), and peak 6 (chenodeoxycholic acid) are more significant than the other peaks with the standard parameter estimate 0.453, -0.166, 0.749, 0.025, respectively. So, compounds cholic acid, taurocholate sodium, and chenodeoxycholic acid might be the major anti-bacterial components in artificial C. bovis. Altogether, this work provides a general model of the combination of UPLC chromatography and anti-bacterial effect to study the spectrum-effect relationships of artificial C. bovis extracts, which can be used to discover the main anti-bacterial components in artificial C. bovis or other Chinese herbal medicines with anti-bacterial effects. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  1. ITER and research works on magnetic confinement fusion

    International Nuclear Information System (INIS)

    Jacquinot, J.

    2013-01-01

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

  2. ICP (ITER Collaborative Platform)

    Energy Technology Data Exchange (ETDEWEB)

    Capuano, C.; Carayon, F.; Patel, V. [ITER, 13 - St. Paul-Lez Durance (France)

    2009-07-01

    The ITER organization has the necessity to manage a massive amount of data and processes. Each team requires different process and databases often interconnected with those of others teams. ICP is the current central ITER repository of structured and unstructured data. All data in ICP is served and managed via a web interface that provides global accessibility with a common user friendly interface. This paper will explain the model used by ICP and how it serves the ITER project by providing a robust and agile platform. ICP is developed in ASP.NET using MSSQL Server for data storage. It currently houses 15 data driven applications, 150 different types of record, 500 k objects and 2.5 M references. During European working hours the system averages 150 concurrent users and 20 requests per second. ICP connects to external database applications to provide a single entry point to ITER data and a safe shared storage place to maintain this data long-term. The Core model provides an easy to extend framework to meet the future needs of the Organization. ICP follows a multi-tier architecture, providing logical separation of process. The standard three-tier architecture is expanded, with the data layer separated into data storage, data structure, and data access components. The business or applications logic layer is broken up into a common business functionality layer, a type specific logic layer, and a detached work-flow layer. Finally the presentation tier comprises a presentation adapter layer and an interface layer. Each layer is built up from small blocks which can be combined to create a wide range of more complex functionality. Each new object type developed gains access to a wealth of existing code functionality, while also free to adapt and extend this. The hardware structure is designed to provide complete redundancy, high availability and to handle high load. This document is composed of an abstract followed by the presentation transparencies. (authors)

  3. Cooperation between CERN and ITER

    CERN Document Server

    2008-01-01

    CERN and the International Fusion Organisation ITER have just signed a first cooperation agreeement. Kaname Ikeda, the Director-General of the International Fusion Energy Organisation (ITER) (on the right) and Robert Aymar, Director-General of CERN, signing the agreement.The Director-General of the International Fusion Energy Organization, Mr Kaname Ikeda, and CERN Director-General, Robert Aymar, signed a cooperation agreement at a meeting on the Meyrin site on Thursday 6 March. One of the main purposes of this agreement is for CERN to give ITER the benefit of its experience in the field of technology as well as in administrative domains such as finance, procurement, human resources and informatics through the provision of consultancy services. Currently in its start-up phase at its Cadarache site, 70 km from Marseilles (France), ITER will focus its research on the scientific and technical feasibility of using fusion energy as a fu...

  4. Analysis of the ITER cryoplant operational modes

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  5. The main component of an alarm pheromone of kissing bugs plays multiple roles in the cognitive modulation of the escape response

    Directory of Open Access Journals (Sweden)

    Sebastián eMinoli

    2013-07-01

    Full Text Available Innate responses in animals can be modulated by experience. Disturbed adults of the triatomine bug Triatoma infestans release an alarm pheromone (AP that elicits an escape response in conspecific larvae. The main component of this AP, the isobutyric acid (IsoAc, alone has already shown to generate an escape response in this species. However, not much is known about the modulation of this behavior by non-associative and associative cognitive processes. We present here evidences of the cognitive capacities of T. infestans larvae in an escape context under different conditioning paradigms, including IsoAc in different roles. We show that: 1 the duration of a pre-exposure to IsoAc plays a main role in determining the type of non-associative learning expressed: short time pre-exposures elicit a sensitization while a longer pre-exposure time triggers a switch from repellence to attractiveness; 2 a simple pre-exposure event is enough to modulate the escape response of larvae to the AP and to its main component: IsoAc; 3 IsoAc and the AP are perceived as different chemical entities; 4 an association between IsoAc and an aversive stimulus can be created under a classical conditioning paradigm; 5 an association between IsoAc and a self-action can be generated under an operant conditioning. These results evince that IsoAc can attain multiple and different cognitive roles in the modulation of the escape response of triatomines and show how cognitive processes can modulate a key behavior for surviving, as it is the escaping response in presence of a potential danger in insects.

  6. 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

  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. 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.

  9. ITER safety and operational scenario

    International Nuclear Information System (INIS)

    Shimomura, Y.; Saji, G.

    1998-01-01

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

  10. Structural analysis of ITER TBM Frame and Dummy TBM

    International Nuclear Information System (INIS)

    Marin, Anna; Kim, Byoung Yoon; Bertolini, Claudio; Lucca, Flavio; Komarov, Victor; Merola, Mario; Giancarli, Luciano; Gicquel, Stefan

    2013-01-01

    One of the main engineering performance goals of ITER is to test and validate design concepts of tritium breeding blankets. To accomplish these goals, three ITER equatorial ports are dedicated to the test of Test Blanket Modules (TBMs) that are mock-ups of tritium breeding blankets. These TBMs, associated with appropriate shield blocks, will also provide the same thermal and nuclear shielding as the main blanket. The main function of TBM Port Plug (PP) is to accommodate TBMs and provide a standardized interface with the vacuum vessel (VV)/port structure. The feasibility of the design concept of the Frame including two Dummy TBMs has been investigated by proposing design improvements of the reference design through an extensive set of thermal, electromagnetic (EM) and stress analyses. As well, the related static strength was evaluated in accordance with the structural design criteria for ITER in-vessel components (SDC-IC). This paper outlines the engineering aspects of the ITER TBM Frame and Dummy TBM and focuses on the feasibility of the present design by structural assessment

  11. Determination of the Three Main Components of the Grapevine Moth Pest Pheromone in Grape-Related Samples by Headspace-Gas Chromatography-Mass Spectrometry

    Directory of Open Access Journals (Sweden)

    María del Carmen Alcudia-León

    2017-10-01

    Full Text Available The grapevine moth (Lobesia botrana is the most significant pest of viticulture. This article reports the development of an analytical method that allows the instrumental determination of the three main pheromone components of the pest ((E,Z-7,9-dodecadien-1-yl acetate, (E,Z-7,9-dodecadien-1-ol and (Z-9-dodecen-1-yl acetate in grape-related samples (must, table grape and wine grape. The combination of headspace, gas chromatography and mass spectrometry provides limits of detection in the range of 60–420 ng/Kg and precision, expressed as a relative standard deviation, better than 8.5%. This analytical approach is rapid and simple and opens a door to the study of the pest incidence on the final products.

  12. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

    Kakudate, S.; Nakahira, M.; Oka, K.; Taguchi, K.; Obara, K.; Tada, E.; Shibanuma, K.; Tesini, A.; Haange, R.; Maisonnier, D.

    2001-01-01

    In ITER, the in-vessel components such as blanket are to be maintained or replaced remotely since they will be activated by 14 MeV neutrons, and a complete exchange of shielding blanket with breeding blanket is foreseen after the Basic Performance Phase. The blanket is segmented into about seven hundred modules to facilitate remote maintainability and allow individual module replacement. For this, the remote handing equipment for blanket maintenance is required to handle a module with a dead weight of about 4 tonne within a positioning accuracy of a few mm under intense gamma radiation. According to the ITER R and D program, a rail-mounted vehicle manipulator system was developed and the basic feasibility of this system was verified through prototype testing. Following this, development of full-scale remote handling equipment has been conducted as one of the ITER Seven R and D Projects aiming at a remote handling demonstration of the ITER blanket. As a result, the Blanket Test Platform (BTP) composed of the full-scale remote handling equipment has been completed and the first integrated performance test in March 1998 has shown that the fabricate remote handling equipment satisfies the main requirements of ITER blanket maintenance. (author)

  13. Radiation sensitivity of microorganisms adhering to the crude drug ''Bezoar Bovis'' and stability of its main components for γ-ray irradiation

    International Nuclear Information System (INIS)

    Ohnishi, Tokuhiro; Okamoto, Shinichi; Kimura, Syojiro; Taimatsu, Meiko; Endo, Akira.

    1994-01-01

    The sterilization dose (SD) of γ rays required for the microbial-contaminated crude drug ''Bezoar Bovis'' and the residual rates of its characteristic and effective components, bilirubin and some kinds of cholic acid, were studied experimentally. Samples of Bezoar Bovis made in America were used. The contamination level of the samples was 2.2 x 10 8 cells of bacteria and 6.0 x 10 5 spores of fungi per g specimen. The survival rate of these microorganisms showed nearly an exponential dependence on radiation dose. The decimal reduction doses (D 10 ) for the bacteria and fungi were found to be 1.5 kGy and 1.1 kGy respectively. From these values, the dose required for attaining the contamination level provided by the administrative guidance (Bacteria, 3 cells/g; Fungi, 2 spores/g) were estimated to be 7.5 kGy and 4.3 kGy, respectively. The G-values for bilirubin, cholic acid and deoxycholic acid were calculated to be 13, 6 and 8, respectively. If the sterilization treatment is carried out with a dose less than 10 kGy on the specimen in a dry powder state, the reduction of the main components such as bilirubin and cholic acids by γ-ray irradiation is considered to be negligible. (author)

  14. Crystallization and preliminary X-ray analysis of PaaAC, the main component of the hydroxylase of the Escherichia coli phenylacetyl-coenzyme A oxygenase complex

    International Nuclear Information System (INIS)

    Grishin, Andrey M.; Ajamian, Eunice; Zhang, Linhua; Cygler, Miroslaw

    2010-01-01

    The expression, purification, crystallization and preliminary crystallographic analysis of the PaaAC complex is reported. This is the main component of the E. coliphenylacetyl-coenzyme A oxygenase complex. The Escherichia coli paa operon encodes enzymes of the phenylacetic acid-utilization pathway that metabolizes phenylacetate in the form of a coenzyme A (CoA) derivative. The phenylacetyl-coenzyme A oxygenase complex, which has been postulated to contain five components designated PaaABCDE, catalyzes ring hydroxylation of phenylacetyl-CoA. The PaaAC subcomplex shows low sequence similarity to other bacterial multicomponent monooxygenases (BMMs) and forms a separate branch on the phylogenetic tree. PaaAC, which catalyzes the hydroxylation reaction, was purified and crystallized in the absence of a bound ligand as well as in complexes with CoA, 3-hydroxybutyryl-CoA, benzoyl-CoA and the true substrate phenylacetyl-CoA. Crystals of the ligand-free enzyme belonged to space group P2 1 2 1 2 1 and diffracted to 2.65 Å resolution, whereas complexes with CoA and its derivatives crystallized in space group P4 1 2 1 2 and diffracted to ∼2.0 Å resolution. PaaAC represents the first crystallized BMM hydroxylase that utilizes a CoA-linked substrate

  15. Iterative Splitting Methods for Differential Equations

    CERN Document Server

    Geiser, Juergen

    2011-01-01

    Iterative Splitting Methods for Differential Equations explains how to solve evolution equations via novel iterative-based splitting methods that efficiently use computational and memory resources. It focuses on systems of parabolic and hyperbolic equations, including convection-diffusion-reaction equations, heat equations, and wave equations. In the theoretical part of the book, the author discusses the main theorems and results of the stability and consistency analysis for ordinary differential equations. He then presents extensions of the iterative splitting methods to partial differential

  16. 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

  17. 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.)

  18. ITER waste management

    International Nuclear Information System (INIS)

    Rosanvallon, S.; Na, B.C.; Benchikhoune, M.; Uzan, J. Elbez; Gastaldi, O.; Taylor, N.; Rodriguez, L.

    2010-01-01

    ITER will produce solid radioactive waste during its operation (arising from the replacement of components and from process and housekeeping waste) and during decommissioning (de-activation phase and dismantling). The waste will be activated by neutrons of energies up to 14 MeV and potentially contaminated by activated corrosion products, activated dust and tritium. This paper describes the waste origin, the waste classification as a function of the French national agency for radioactive waste management (ANDRA), the optimization process put in place to reduce the waste radiotoxicity and volumes, the estimated waste amount based on the current design and maintenance procedure, and the overall strategy from component removal to final disposal anticipated at this stage of the project.

  19. ITER council proceedings: 1998

    International Nuclear Information System (INIS)

    1999-01-01

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

  20. Comparison of thermal behavior of natural and hot-washed sisal fibers based on their main components: Cellulose, xylan and lignin. TG-FTIR analysis of volatile products

    Energy Technology Data Exchange (ETDEWEB)

    Benítez-Guerrero, Mónica, E-mail: monica_benitez_guerrero@yahoo.es [Departamento de Ingeniería Civil, Materiales y Fabricación, Universidad de Málaga, Escuela de Ingenierías, C/ Dr. Ortiz Ramos s/n, Campus Teatinos, 29071 Málaga (Spain); López-Beceiro, Jorge [Departamento de Ingeniería Industrial II, Escola Politécnica Superior, Universidade da Coruña, Avda. Mendizábal, 15403 Ferrol (Spain); Sánchez-Jiménez, Pedro E. [Instituto de Ciencia de Materiales de Sevilla, CSIC-Universidad de Sevilla, C/ Américo Vespucio 49, 41092 Sevilla (Spain); Pascual-Cosp, José [Departamento de Ingeniería Civil, Materiales y Fabricación, Universidad de Málaga, Escuela de Ingenierías, C/ Dr. Ortiz Ramos s/n, Campus Teatinos, 29071 Málaga (Spain)

    2014-04-01

    Highlights: • Thermal decomposition of sisal fibers has been discussed. • Decompositions of lignocellulosic components and sisal are compared by TXRD and TG-FTIR. • Hot washing reduces the temperature range in which sisal decomposition occurs. • Sisal cellulose decomposition goes by an alternative route to levoglucosan generation. - Abstract: This paper presents in a comprehensive way the thermal behavior of natural and hot-washed sisal fibers, based on the fundamental components of lignocellulosic materials: cellulose, xylan and lignin. The research highlights the influence exerted on the thermal stability of sisal fibers by other constituents such as non-cellulosic polysaccharides (NCP) and mineral matter. Thermal changes were investigated by thermal X-ray diffraction (TXRD), analyzing the crystallinity index (%Ic) of cellulosic samples, and by simultaneous thermogravimetric and differential thermal analysis coupled with Fourier-transformed infrared spectrometry (TG/DTA-FTIR), which allowed to examine the evolution of the main volatile compounds evolved during the degradation under inert and oxidizing atmospheres. The work demonstrates the potential of this technique to elucidate different steps during the thermal decomposition of sisal, providing extensible results to other lignocellulosic fibers, through the analysis of the evolution of CO{sub 2}, CO, H{sub 2}O, CH{sub 4}, acetic acid, formic acid, methanol, formaldehyde and 2-butanone, and comparing it with the volatile products from pyrolysis of the biomass components. The hydroxyacetaldehyde detected during pyrolysis of sisal is indicative of an alternative route to that of levoglucosan, generated during cellulose pyrolysis. Hot-washing at 75 °C mostly extracts non-cellulosic components of low decomposition temperature, and reduces the range of temperature in which sisal decomposition occurs, causing a retard in the pyrolysis stage and increasing Tb{sub NCP} and Tb{sub CEL}, temperatures at the

  1. The WEST project: Current status of the ITER-like tungsten divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Bucalossi, J.; Corre, Y.; Ferlay, F.; Firdaouss, M.; Garin, P.; Grosman, A.; Guilhem, D.; Gunn, J.; Languille, P.; Lipa, M.; Richou, M.; Tsitrone, E.

    2014-01-01

    Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues

  2. The WEST project: Current status of the ITER-like tungsten divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M., E-mail: marc.missirlian@cea.fr; Bucalossi, J.; Corre, Y.; Ferlay, F.; Firdaouss, M.; Garin, P.; Grosman, A.; Guilhem, D.; Gunn, J.; Languille, P.; Lipa, M.; Richou, M.; Tsitrone, E.

    2014-10-15

    Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.

  3. ITER Operating Limits and Conditions

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  4. Existence test for asynchronous interval iterations

    DEFF Research Database (Denmark)

    Madsen, Kaj; Caprani, O.; Stauning, Ole

    1997-01-01

    In the search for regions that contain fixed points ofa real function of several variables, tests based on interval calculationscan be used to establish existence ornon-existence of fixed points in regions that are examined in the course ofthe search. The search can e.g. be performed...... as a synchronous (sequential) interval iteration:In each iteration step all components of the iterate are calculatedbased on the previous iterate. In this case it is straight forward to base simple interval existence and non-existencetests on the calculations done in each step of the iteration. The search can also...... on thecomponentwise calculations done in the course of the iteration. These componentwisetests are useful for parallel implementation of the search, sincethe tests can then be performed local to each processor and only when a test issuccessful do a processor communicate this result to other processors....

  5. 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

  6. Design analysis of the ITER divertor

    International Nuclear Information System (INIS)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F.; Merola, M.; Riccardi, B.; Petrizzi, L.; Villari, R.

    2007-01-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  7. Design analysis of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F. [L.T. Calcoli SaS, Merate (Lecco) (Italy); Merola, M. [ITER Team, Cadarache (France); Riccardi, B. [EFDA CSU Garching (Germany); Petrizzi, L.; Villari, R. [CRE ENEA sulla Fusione Frascati, Roma (Italy)

    2007-07-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  8. ITER Council proceedings: 1993

    International Nuclear Information System (INIS)

    1994-01-01

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

  9. ITER council proceedings: 2000

    International Nuclear Information System (INIS)

    2001-01-01

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

  10. ITER council proceedings: 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

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

  11. Status of the ITER EDA

    International Nuclear Information System (INIS)

    Aymar, R.

    1999-01-01

    This article summarises progress made in the ITER Design Activities between October 1998 and February 1999. The three main focusses of the activity were on design work, on R and D work and on the physics basis. The consequences of diminishing financial funds and personnel are discussed and the state of the individual R and D projects is given briefly

  12. Informal meeting on ITER developments

    International Nuclear Information System (INIS)

    Canobbio, E.

    2000-01-01

    The International Fusion Research Council (IFRC), advisory body of the IAEA, organized an informal meeting on the general status and outlook for ITER, held October 9 at Sorrento, Italy, in conjunction with the 18th IAEA Fusion Energy Conference. This article describes the main events at the meeting

  13. Third ITER International Industry Liaison Meeting

    International Nuclear Information System (INIS)

    Dautovich, D.

    2000-01-01

    Following previous meetings held in 1996 in San Diego and in 1997 in Tokyo, the Third ITER International Industry Liaison Meeting (IILM) meeting was held under the European Chairmanship in Toronto, Canada, November 7-9, 2000. The intention of such meetings is to provide a forum for industrialists of the ITER EDA parties and other interested countries to develop common understandings on important issues of the timing and nature of Industry involvement in the ITER project. This article describes the main views from Industry on the preconstruction and construction phases and the cost and benefit schemes, while summarizing the progress made by the ITER project since the Tokyo meeting

  14. ITER Central Solenoid Module Fabrication

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-23

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

  15. Assembly process of the ITER neutral beam injectors

    Energy Technology Data Exchange (ETDEWEB)

    Graceffa, J., E-mail: joseph.graceffa@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul lez Durance (France); Boilson, D.; Hemsworth, R.; Petrov, V.; Schunke, B.; Urbani, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul lez Durance (France); Pilard, V. [Fusion for Energy, C/ Josep Pla, n°2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2013-10-15

    The ITER neutral beam (NB) injectors are used for heating and diagnostics operations. There are 4 injectors in total, 3 heating neutral beam injectors (HNBs) and one diagnostic neutral beam injector (DNB). Two HNBs and the DNB will start injection into ITER during the hydrogen/helium phase of ITER operations. A third HNB is considered as an upgrade to the ITER heating systems, and the impact of the later installation and use of that injector have to be taken into account when considering the installation and assembly of the whole NB system. It is assumed that if a third HNB is to be installed, it will be installed before the nuclear phase of the ITER project. The total weight of one injector is around 1200 t and it is composed of 18 main components and 36 sets of shielding plates. The overall dimensions are length 20 m, height 10 m and width 5 m. Assembly of the first two HNBs and the DNB will start before the first plasma is produced in ITER, but as the time required to assemble one injector is estimated at around 1.5 year, the assembly will be divided into 2 steps, one prior to first plasma, and the second during the machine second assembly phase. To comply with this challenging schedule the assembly sequence has been defined to allow assembly of three first injectors in parallel. Due to the similar design between the DNB and HNBs it has been decided to use the same tools, which will be designed to accommodate the differences between the two sets of components. This reduces the global cost of the assembly and the overall assembly time for the injector system. The alignment and positioning of the injectors is a major consideration for the injector assembly as the alignment of the beamline components and the beam source are critical if good injector performance is to be achieved. The theoretical axes of the beams are defined relative to the duct liners which are installed in the NB ports. The concept adopted to achieve the required alignment accuracy is to use the

  16. Influence of Laser Radiation Power Density on the Intensity of Spectral Lines for Main Components in a Clay Laser-Induced Plasma

    Science.gov (United States)

    Anufrik, S. S.; Kurian, N. N.; Znosko, K. F.; Belkov, M. V.

    2018-05-01

    We have studied the intensity of the spectral lines for the main components in clay: Al I 309.4 nm, Al II 358.7 nm, Mg II 279.6 nm, Ti II 323.6 nm vs. the position of the object relative to the focus of the optical system when the samples are exposed to single laser pulses from a YAG:Nd3+ laser. We have determined the permissible ranges for positioning the object relative to the focus of the optical system (positive and negative defocusing) for which there is practically no change in the reproducibility of the intensity for the spectral lines for red and white clay samples. We show that the position of the object relative to the focus of the optical system should be within the range ΔZ ±1.5 mm for optimal laser pulse energies for the analyte spectral lines. We have calculated the radiation flux density for different laser pulse energies and different distances from the focus to the object. We have shown experimentally that reducing the radiation flux density leads to a decrease in the intensity of the analyte spectral lines.

  17. Evaluation of the Use of Different Solvents to Extract the Four Main Components of Yinqiaosan and Their In Vitro Inhibitory Effects on Influenza-A Virus

    Directory of Open Access Journals (Sweden)

    Xuefeng Wang

    2010-04-01

    Full Text Available High-performance liquid chromatography coupled with an ultraviolet detector was used to quantitatively determine the amount of caffeoylquinic acid, arctium lappa glycoside, aurantiamarin and forsythin present in five solvent extracts (water, 30% ethanol, 50% ethanol, 70% ethanol and 95% ethanol of Yinqiaosan (YQS. The calibration curves of the four components were linear (r2 > 0.998 over the concentration range of 0.9–151.5 μg/mL. This method showed good repeatability with relative standard deviations of less than 2.9% and mean recoveries of 91.3–100.5%. We established the relative amounts of each of the four main ingredients in YQS for all five solvent extracts. The direct inhibitory effects of the five YQS extracts on influenza-A virus in vitro and the therapeutic effects on infected embryonated hen eggs were evaluated. YQS extracted in 50% ethanol and used at concentrations of 500mg/mL and 250mg/mL exerted the greatest therapeutic effect in infected embryonated hen eggs and was significantly superior to ribavirin (p < 0.05. Therefore, the 50% ethanol extract of YQS seems to be the most effective for the inhibition of influenza-A virus. Based on these results, further studies are warranted to evaluate the use of YQS extracts as a possible therapy for influenza.

  18. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B.; Janeschitz, G. [Forschungszentrum Karlsruhe GmbH, FZK, Karlsruhe (Germany); Landman, I.; Pestchanyi, S. [FZK-Forschungszentrum Karlsruhe, Association Euratom-FZK, Technik und Umwelt, Karlsruhe (Germany); Loarte, A. [EFDA Close Support Unit Garching, Garching bei Munchen(Germany)

    2007-07-01

    Full text of publication follows: Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient release of energy from the confined plasma onto divertor and the first wall by multiple ELMs (about 10{sup 4} ELMs per ITER discharge), which can play a determining role in the erosion rate and lifetime of these components. It is expected that about 50-70 % of the ELM energy releases onto divertor armour and the rest is dumped onto the First Wall (FW) armour. The expected energy heat loads on the ITER divertor and FW during Type I ELM are in range 0.5 - 4 MJ/m{sup 2} in timescales of 0.3-0.6 ms. In case of the ITER disruptions the material evaporated from the divertor expands into the SOL and generates significant radiation heating of the FW armour up to several GW/m2 during a few milliseconds that can also lead to the its melting and noticeable damage. Beryllium macro-brush armour (Be-brushes) is foreseen as plasma FW facing component (PFC) in ITER. During the intense transient events in ITER the surface melting, melt motion, melt splashing and evaporation are seen as the main mechanisms of Be-erosion. The expected erosion of the ITER plasma facing components under transient energy loads can be properly estimated by numerical simulations using the codes MEMOS and PHEMOBRID validated against experimental data obtained at the plasma gun facilities QSPA-T, MK-200UG and QSPA-Kh50 that provide a way to simulate the energy loads expected in ITER in laboratory experiments. The numerical simulations were carried out for the expected ITER ELMs for the heat loads in the range 0.5 - 3.0 MJ/m{sup 2} and the timescale up 0.6 ms and ITER disruptions for the heat loads in the range 2 - 13 MJ/m{sup 2} in timescales of 1-5 ms. Radiation heat loads at the FW armour from the vapour expanded into the SOL were calculated using the codes FOREV-2 and TOKES for both ITER ELM and ITER disruption scenarios. Melt layer damage of the Be

  19. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    International Nuclear Information System (INIS)

    Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2007-01-01

    Full text of publication follows: Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient release of energy from the confined plasma onto divertor and the first wall by multiple ELMs (about 10 4 ELMs per ITER discharge), which can play a determining role in the erosion rate and lifetime of these components. It is expected that about 50-70 % of the ELM energy releases onto divertor armour and the rest is dumped onto the First Wall (FW) armour. The expected energy heat loads on the ITER divertor and FW during Type I ELM are in range 0.5 - 4 MJ/m 2 in timescales of 0.3-0.6 ms. In case of the ITER disruptions the material evaporated from the divertor expands into the SOL and generates significant radiation heating of the FW armour up to several GW/m2 during a few milliseconds that can also lead to the its melting and noticeable damage. Beryllium macro-brush armour (Be-brushes) is foreseen as plasma FW facing component (PFC) in ITER. During the intense transient events in ITER the surface melting, melt motion, melt splashing and evaporation are seen as the main mechanisms of Be-erosion. The expected erosion of the ITER plasma facing components under transient energy loads can be properly estimated by numerical simulations using the codes MEMOS and PHEMOBRID validated against experimental data obtained at the plasma gun facilities QSPA-T, MK-200UG and QSPA-Kh50 that provide a way to simulate the energy loads expected in ITER in laboratory experiments. The numerical simulations were carried out for the expected ITER ELMs for the heat loads in the range 0.5 - 3.0 MJ/m 2 and the timescale up 0.6 ms and ITER disruptions for the heat loads in the range 2 - 13 MJ/m 2 in timescales of 1-5 ms. Radiation heat loads at the FW armour from the vapour expanded into the SOL were calculated using the codes FOREV-2 and TOKES for both ITER ELM and ITER disruption scenarios. Melt layer damage of the Be FW macro

  20. 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)

  1. ITER council proceedings: 1995

    International Nuclear Information System (INIS)

    1996-01-01

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

  2. 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.

  3. Characteristics and environmental fate of the anionic surfactant sodium lauryl ether sulphate (SLES) used as the main component in foaming agents for mechanized tunnelling

    International Nuclear Information System (INIS)

    Barra Caracciolo, Anna; Cardoni, Martina; Pescatore, Tanita; Patrolecco, Luisa

    2017-01-01

    The anionic surfactant sodium lauryl ether sulphate (SLES) is the main component of most commercial products used for soil conditioning in the excavation industry, in particular as lubricants for mechanized tunnelling. Its use during the excavation processes can result in either the subsequent possible re-use of the huge amount of soil debris as by-products (e.g. land covering) or its discharge as waste. Currently, there are neither SLES soil threshold limits in European legislation, nor comprehensive studies on the environmental risk for soil ecosystems in these exposure scenarios. In this context, the present paper reviews the available data on the intrinsic characteristics of persistence and the ecotoxicological effects of the anionic surfactant SLES. Although SLES is generally reported to be biodegradable in standard tests, with degradation rates between 7 h and 30 days, depending on the initial conditions, data on its biodegradation in environmental studies are quite scarce. Consequently, assessing SLES biodegradation rates in field conditions is crucial for evaluating if in residual concentrations (typically in the range 40–500 mg/kg in excavated soils) it can or not be a potential hazard for terrestrial and water organisms. Laboratory ecotoxicological tests pointed out detrimental effects of SLES for aquatic organisms, while data on the terrestrial species are rather poor so far and further studies at the expected environmental concentrations are necessary. Finally, the review reports the main analytical methods available for detecting anionic surfactants in solid matrices and the future research needed to improve knowledge on the possible environmental risks posed by the use of SLES in foaming agents for mechanized tunnelling. - Highlights: • Tons of excavated soil containing SLES are reused posing an environmental risk. • SLES can have detrimental effects on aquatic organisms exposed in lab test. • There is a need to improve knowledge on SLES

  4. ITER council proceedings: 1999

    International Nuclear Information System (INIS)

    1999-01-01

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

  5. ITER council proceedings: 1996

    International Nuclear Information System (INIS)

    1997-01-01

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

  6. ITER EDA technical activities

    International Nuclear Information System (INIS)

    Aymar, R.

    1998-01-01

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

  7. 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

  8. Effects of clary sage oil and its main components, linalool and linalyl acetate, on the plasma membrane of Candida albicans: an in vivo EPR study.

    Science.gov (United States)

    Blaskó, Ágnes; Gazdag, Zoltán; Gróf, Pál; Máté, Gábor; Sárosi, Szilvia; Krisch, Judit; Vágvölgyi, Csaba; Makszin, Lilla; Pesti, Miklós

    2017-02-01

    The effects of clary sage (Salvia sclarea L.) oil (CS-oil), and its two main components, linalool (Lol) and linalyl acetate (LA), on cells of the eukaryotic human pathogen yeast Candida albicans were studied. Dynamic and thermodynamic properties of the plasma membrane were investigated by electron paramagnetic resonance (EPR) spectroscopy, with 5-doxylstearic acid (5-SASL) and 16-SASL as spin labels. The monitoring of the head group regions with 5-SASL revealed break-point frequency decrease in a temperature dependent manner of the plasma membrane between 9.55 and 13.15 °C in untreated, in CS-oil-, Lol- and LA-treated membranes. The results suggest a significant increase in fluidity of the treated plasma membranes close to the head groups. Comparison of the results observed with the two spin labels demonstrated that CS-oil and LA induced an increased level of fluidization at both depths of the plasma membrane. Whereas Lol treatment induced a less (1 %) ordered bilayer organization in the superficial regions and an increased (10 %) order of the membrane leaflet in deeper layers. Acute toxicity tests and EPR results indicated that both the apoptotic and the effects exerted on the plasma membrane fluidity depended on the composition and chemical structure of the examined materials. In comparison with the control, treatment with CS-oil, Lol or LA induced 13.0, 12.3 and 26.4 % loss respectively, of the metabolites absorbing at 260 nm, as a biological consequence of the plasma membrane fluidizing effects. Our results confirmed that clary sage oil causes plasma membrane perturbations which leads to cell apoptosis process.

  9. Remote handling maintenance of ITER

    International Nuclear Information System (INIS)

    Haange, R.

    1999-01-01

    The remote maintenance strategy and the associated component design of the International Thermonuclear Experimental Reactor (ITER) have reached a high degree of completeness, especially with respect to those components that are expected to require frequent or occasional remote maintenance. Large-scale test stands, to demonstrate the principle feasibility of the remote maintenance procedures and to develop the required equipment and tools, were operational at the end of the Engineering Design Activities (EDA) phase. The initial results are highly encouraging: major remote equipment deployment and component replacement operations have been successfully demonstrated. (author)

  10. ITER-FEAT safety

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  11. ITER council proceedings: 1997

    International Nuclear Information System (INIS)

    1997-01-01

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

  12. Studies for site preparation for ITER construction

    International Nuclear Information System (INIS)

    Fardeau, A.; Blanc, F.; Cardettini, J.D.; Mandine, J.R.; Guerin, R.; Patisson, L.; Bergegere, P.; Santagiustina, A.; Garin, P.

    2004-01-01

    The implantation of a nuclear facility such as ITER (about 20 buildings on 40 ha) requires many preparatory studies, particularly with respect to: Underground characteristics; Topography, layout; Deforestation, excavations; Networks, fences and roads; Impact of seismic hazard on design. This paper presents the main results of these studies, carried out within the European ITER Site Studies framework. A dedicated paper in the conference deals with the transport studies. To choose the site for ITER implantation, detailed geological, hydrogeological and geophysical investigations have been carried out. Taking into account the meteorological data (particularly the main wind direction), topography, access, electrical supply, fluids needs and constraints (gravity systems), buildings and roads have been implemented. (authors)

  13. 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.

  14. International Workshop on Diagnostics for ITER

    CERN Document Server

    Gorini, Giuseppe; Sindoni, Elio

    1996-01-01

    This book of proceedings collects the papers presented at the Workshop on Diagnostics for ITER, held at Villa Monastero, Varenna (Italy), from August 28 to September 1, 1995. The Workshop was organised by the International School of Plasma Physics "Piero Caldirola. " Established in 1971, the ISPP has organised over fifty advanced courses and workshops on topics mainly related to plasma physics. In particular, courses and workshops on plasma diagnostics (previously held in 1975, 1978, 1982, 1986, and 1991) can be considered milestones in the history of this institution. Looking back at the proceedings of the previous meetings in Varenna, one can appreciate the rapid progress in the field of plasma diagnostics over the past 20 years. The 1995 workshop was co-organised by the Istituto di Fisica del Plasma of the National Research Council (CNR). In contrast to previous Varenna meetings on diagnostics, which have covered diagnostics in present-day tokamaks and which have had a substantial tutorial component, the 1...

  15. Critical Assessment of Pressure Gauges for ITER

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  16. Disruption modeling in support of ITER

    International Nuclear Information System (INIS)

    Bandyopadhyay, I.

    2015-01-01

    Plasma current disruptions and Vertical Displacement Events (VDEs) are one of the major concerns in any tokamak as they lead to large electromagnetic forces to tokamak first wall components and vacuum vessel. Their occurrence also means disruption to steady state operations of tokamaks. Thus future fusion reactors like ITER must ensure that disruptions and VDEs are minimized. However, since there is still finite probability of their occurrence, one must be able to characterize disruptions and VDEs and able to predict, for example, the plasma current quench time and halo current amplitude, which mainly determine the magnitude of the electromagnetic forces. There is a concerted effort globally to understand and predict plasma and halo current evolution during disruption in tokamaks through MHD simulations. Even though Disruption and VDEs are often 3D MHD perturbations in nature, presently they are mostly simulated using 2D axisymmetric MHD codes like the Tokamak Simulation Code (TSC) and DINA. These codes are also extensively benchmarked against experimental data in present day tokamaks to improve these models and their ability to predict these events in ITER. More detailed 3D models like M3D are only recently being developed, but they are yet to be benchmarked against experiments, as also they are massively computationally exhaustive

  17. Starting manufacturing phase of ITER upper ports

    Energy Technology Data Exchange (ETDEWEB)

    Utin, Yuri, E-mail: yuri.utin@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Alekseev, Alexander; Sborchia, Carlo; Choi, Changho; Albin, Vincent; Barabash, Vladimir; Davis, James [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Fabritsiev, Sergey [NTC Sintez, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Giraud, Benoit; Guirao, Julio [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Koenig, Werner [MAN Diesel & Turbo SE, Werftstrasse 17, Deggendorf (Germany); Kedrov, Igor; Kuzmin, Evgeny [NTC Sintez, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Levesy, Bruno; Martinez, Jean-Marc [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Prebeck, Markus [MAN Diesel & Turbo SE, Werftstrasse 17, Deggendorf (Germany); Privalova, Elena [NTC Sintez, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Ranzinger, Franz [MAN Diesel & Turbo SE, Werftstrasse 17, Deggendorf (Germany); Savrukhin, Petr [Russian Federation ITER Domestic Agency, Kurchatov sq.1, 123182 Moscow (Russian Federation); Schiller, Thomas [MAN Diesel & Turbo SE, Werftstrasse 17, Deggendorf (Germany); and others

    2015-10-15

    Highlights: • The port plugs are attached to the ports with high-strength fasteners. • Tightening of the fasteners via inductive heating was tested. • A concept for the port/plug sealing with metal-type gaskets has progressed. • Manufacturing design of the Upper Ports is in progress. • A full-scale mock-up of double-wall part of the port stub extension is in manufacturing process – acceptable final tolerances are expected. - Abstract: The ITER Vacuum Vessel (VV) features upper, equatorial and lower ports. The upper and regular equatorial ports are occupied by the port plugs. Although the port design has been overall completed in the past, the design of some remaining interfaces was still in progress: in particular, the Sealing Flange package, which includes the high-vacuum seals and the plug fasteners. As the ITER construction phase has started, the procurement of the VV ports has been launched. The VV upper ports will be procured by the Russian Federation Domestic Agency. The main suppliers were selected and the manufacturing design of the first parts is in full progress now. Since the VV is classified at nuclear level N2, the design and manufacture of its components are to be compliant with the French RCC-MR code and regulations for nuclear pressure equipment in France. These regulations make a strong impact to the port design and manufacturing process.

  18. Starting manufacturing phase of ITER upper ports

    International Nuclear Information System (INIS)

    Utin, Yuri; Alekseev, Alexander; Sborchia, Carlo; Choi, Changho; Albin, Vincent; Barabash, Vladimir; Davis, James; Fabritsiev, Sergey; Giraud, Benoit; Guirao, Julio; Koenig, Werner; Kedrov, Igor; Kuzmin, Evgeny; Levesy, Bruno; Martinez, Jean-Marc; Prebeck, Markus; Privalova, Elena; Ranzinger, Franz; Savrukhin, Petr; Schiller, Thomas

    2015-01-01

    Highlights: • The port plugs are attached to the ports with high-strength fasteners. • Tightening of the fasteners via inductive heating was tested. • A concept for the port/plug sealing with metal-type gaskets has progressed. • Manufacturing design of the Upper Ports is in progress. • A full-scale mock-up of double-wall part of the port stub extension is in manufacturing process – acceptable final tolerances are expected. - Abstract: The ITER Vacuum Vessel (VV) features upper, equatorial and lower ports. The upper and regular equatorial ports are occupied by the port plugs. Although the port design has been overall completed in the past, the design of some remaining interfaces was still in progress: in particular, the Sealing Flange package, which includes the high-vacuum seals and the plug fasteners. As the ITER construction phase has started, the procurement of the VV ports has been launched. The VV upper ports will be procured by the Russian Federation Domestic Agency. The main suppliers were selected and the manufacturing design of the first parts is in full progress now. Since the VV is classified at nuclear level N2, the design and manufacture of its components are to be compliant with the French RCC-MR code and regulations for nuclear pressure equipment in France. These regulations make a strong impact to the port design and manufacturing process.

  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. Status of international collaborative efforts on selected ITER materials

    International Nuclear Information System (INIS)

    Belyakov, V.A.; Fabritsiev, S.A.; Mazul, I.V.; Rowcliffe, A.F.

    2000-01-01

    The paper presents an overview of the performance of refractory metals, beryllium, and copper alloys, for the international thermonuclear experimental reactor (ITER) high heat flux structures. High temperature brazing, hot isostatic pressing (HIP), friction welding, explosive bonding, and other methods were explored to join copper alloys to 316 stainless steel for first wall and limiter applications. It is concluded that the main material problems for the ITER high heat flux components are: (a) degradation of properties after the manufacturing cycle (especially for Be/Cu and Cu/stainless steel (SS) joints); (b) helium embrittlement of Be, and Cu, and; (c) radiation-induced loss of fracture toughness for Be, W, and Cu alloys

  1. 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)

  2. 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)

  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. ITER parametric analysis and operational performance

    International Nuclear Information System (INIS)

    Perkins, L.J.; Spears, W.R.; Galambos, J.D.

    1991-01-01

    One of the key components of the ITER Conceptual Design Activities (CDA) is the determination of optimum design, investigation of operation in various modes, recommendation of baseline performance specifications, studies of sensitivity of ITER design to uncertainties in physics, investigation of operational flexibility, assessment of alternative designs, and determination of implications for extrapolation to prospective DEMO reactors. These terms of reference are reported in this document. Refs, figs and tabs

  5. Japanese site for ITER: Rokkasho

    International Nuclear Information System (INIS)

    Kishimoto, Hiroshi

    2003-01-01

    This paper describes the status of Japanese efforts for hosting ITER in Japan. In May 2002, Japanese Government decided to propose an ITER site, Rokkasho in Aomori Prefecture, a Northern part of the main island, based on the comprehensive/intensive assessments by the Site Selection Committee established by Japanese Government. ITER is designed basically with a potential flexibility beyond the detailed technical objectives to have more clear scope for developing technical key elements in a future power plant. Various flexibilities in the construction, operation and decommissioning of ITER are totally assessed. Consequently the Japanese site has been chosen and it satisfies sufficiently not only the Site Requirements and the Site Design Assumptions but also the further extension and flexibilities. In particular the potential for more flexible construction schedule and operations is technically described as well as the fulfillment of the site requirements and its assumptions as the minimum requirements in this paper. The socio-cultural environment is also described briefly because of a key aspect for the scientists and engineers who will participate in the project

  6. Iteration and accelerator dynamics

    International Nuclear Information System (INIS)

    Peggs, S.

    1987-10-01

    Four examples of iteration in accelerator dynamics are studied in this paper. The first three show how iterations of the simplest maps reproduce most of the significant nonlinear behavior in real accelerators. Each of these examples can be easily reproduced by the reader, at the minimal cost of writing only 20 or 40 lines of code. The fourth example outlines a general way to iteratively solve nonlinear difference equations, analytically or numerically

  7. Future plan of ITER

    International Nuclear Information System (INIS)

    Kitsunezaki, Akio

    1998-01-01

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

  8. ITER test programme

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  9. Integrating supervision, control and data acquisition—The ITER Neutral Beam Test Facility experience

    Energy Technology Data Exchange (ETDEWEB)

    Luchetta, A., E-mail: adriano.luchetta@igi.cnr.it; Manduchi, G.; Taliercio, C.; Breda, M.; Capobianco, R.; Molon, F.; Moressa, M.; Simionato, P.; Zampiva, E.

    2016-11-15

    Highlights: • The paper describes the experience gained in the integration of different systems for the control and data acquisition system of the ITER Neutral Beam Test Facility. • It describes the way the different frameworks have been integrated. • It reports some lessons learnt during system integration. • It reports some authors’ considerations about the development the ITER CODAC. - Abstract: The ITER Neutral Beam (NBI) Test Facility, under construction in Padova, Italy consists in the ITER full scale ion source for the heating neutral beam injector, referred to as SPIDER, and the full size prototype injector, referred to as MITICA. The Control and Data Acquisition System (CODAS) for SPIDER has been developed and is going to be in operation in 2016. The system is composed of four main components: Supervision, Slow Control, Fast Control and Data Acquisition. These components interact with each other to carry out the system operation and, since they represent a common pattern in fusion experiments, software frameworks have been used for each (set of) component. In order to reuse as far as possible the architecture developed for SPIDER, it is important to clearly define the boundaries and the interfaces among the system components so that the implementation of any component can be replaced without affecting the overall architecture. This work reports the experience gained in the development of SPIDER components, highlighting the importance in the definition of generic interfaces among component, showing how the specific solutions have been adapted to such interfaces and suggesting possible approaches for the development of other ITER subsystems.

  10. Integrating supervision, control and data acquisition—The ITER Neutral Beam Test Facility experience

    International Nuclear Information System (INIS)

    Luchetta, A.; Manduchi, G.; Taliercio, C.; Breda, M.; Capobianco, R.; Molon, F.; Moressa, M.; Simionato, P.; Zampiva, E.

    2016-01-01

    Highlights: • The paper describes the experience gained in the integration of different systems for the control and data acquisition system of the ITER Neutral Beam Test Facility. • It describes the way the different frameworks have been integrated. • It reports some lessons learnt during system integration. • It reports some authors’ considerations about the development the ITER CODAC. - Abstract: The ITER Neutral Beam (NBI) Test Facility, under construction in Padova, Italy consists in the ITER full scale ion source for the heating neutral beam injector, referred to as SPIDER, and the full size prototype injector, referred to as MITICA. The Control and Data Acquisition System (CODAS) for SPIDER has been developed and is going to be in operation in 2016. The system is composed of four main components: Supervision, Slow Control, Fast Control and Data Acquisition. These components interact with each other to carry out the system operation and, since they represent a common pattern in fusion experiments, software frameworks have been used for each (set of) component. In order to reuse as far as possible the architecture developed for SPIDER, it is important to clearly define the boundaries and the interfaces among the system components so that the implementation of any component can be replaced without affecting the overall architecture. This work reports the experience gained in the development of SPIDER components, highlighting the importance in the definition of generic interfaces among component, showing how the specific solutions have been adapted to such interfaces and suggesting possible approaches for the development of other ITER subsystems.

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

    International Nuclear Information System (INIS)

    Yuan Tao; Feng Kaiming; Chen Zhi; Wang Xiaoyu

    2007-01-01

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

  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. GCS component development cycle

    Science.gov (United States)

    Rodríguez, Jose A.; Macias, Rosa; Molgo, Jordi; Guerra, Dailos; Pi, Marti

    2012-09-01

    The GTC1 is an optical-infrared 10-meter segmented mirror telescope at the ORM observatory in Canary Islands (Spain). First light was at 13/07/2007 and since them it is in the operation phase. The GTC control system (GCS) is a distributed object & component oriented system based on RT-CORBA8 and it is responsible for the management and operation of the telescope, including its instrumentation. GCS has used the Rational Unified process (RUP9) in its development. RUP is an iterative software development process framework. After analysing (use cases) and designing (UML10) any of GCS subsystems, an initial component description of its interface is obtained and from that information a component specification is written. In order to improve the code productivity, GCS has adopted the code generation to transform this component specification into the skeleton of component classes based on a software framework, called Device Component Framework. Using the GCS development tools, based on javadoc and gcc, in only one step, the component is generated, compiled and deployed to be tested for the first time through our GUI inspector. The main advantages of this approach are the following: It reduces the learning curve of new developers and the development error rate, allows a systematic use of design patterns in the development and software reuse, speeds up the deliverables of the software product and massively increase the timescale, design consistency and design quality, and eliminates the future refactoring process required for the code.

  15. United States rejoin ITER

    International Nuclear Information System (INIS)

    Roberts, M.

    2003-01-01

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

  16. ITER CTA newsletter. No. 3

    International Nuclear Information System (INIS)

    2001-11-01

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

  17. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  18. ITER council proceedings: 1992

    International Nuclear Information System (INIS)

    1994-01-01

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

  19. ITER containment structures

    International Nuclear Information System (INIS)

    Sadakov, S.; Fauser, F.; Nelson, B.

    1991-01-01

    This document describes the results and recommendations of the Containment Structures Design Unit (CSDU) on the containment structures for ITER, made in the context of the Conceptual Design Phase. The document describes the following subsystems: (1) the primary vacuum vessel (VV), (2) the attaching locks (AL) of the invessel components, (3) the plasma passive and active stabilizers, (4) the cryostat vessel, and (5) the machine gravity supports. Although for most components reference designs were selected, for some of these alternative design options were described, because unresolved problems necessitate further research and development. Conclusions and future needs are summarized for each of the above subsystems: (1) a reference VV design was selected, while most critical VV future needs are the feasibility studies of manufacturing, assembly, and the repair/disassembly/reassembly by remote handling. Alternative, thin-wall options appear attractive and should be studied further during the Engineering Design Activities; (2) no reference design solution was selected for the AL system, as AL design requirements are extremely difficult and internally contradictory, while there is no existing tokamak precedent, but instead, five different approaches will be further researched early in the Engineering Design Phase; (3) significant progress is reported on passive loops, for which the ''twin-loops'' concept is ready to be advanced into the Engineering Design Phase, and on active coils, where a new coil positioning prevents interference with the blanket removal paths, and the current joints are located in a secondary vacuum or in the atmosphere of the reactor hall, repairable by remote handling; (4) a full metallic welded cryostat design with increased toroidal resistance was chosen, but with a design based on concrete with a thin inner metallic liner as a back-up in case detailed nuclear shielding requirements would force the cryostat to act as biological shield; (5) out

  20. IN-VITRO evidence for the protective properties of the main components of the Mediterranean diet against colorectal cancer: A systematic review.

    Science.gov (United States)

    Rotelli, M T; Bocale, D; De Fazio, M; Ancona, P; Scalera, I; Memeo, R; Travaglio, E; Zbar, A P; Altomare, D F

    2015-09-01

    Epidemiological studies have shown that the incidence and mortality rates of colorectal cancer (CRC) vary over 10-fold worldwide where within Westernized societies lower rates are observed amongst populations living within the Mediterranean basin, suggesting a significant influence of environment and dietary style in CRC carcinogenesis. Interpretation of the data concerning the benefits of mediterranean (MD) diet is difficult in vivo because of the variability of alimentary regimens used, the differing compliance with dietary supplementation and because of the non-uniform duration of patient cohort observation. Therefore, the aim of this review is to evaluate the in-vitro effects on colorectal cancer cell lines. the literature concerning the in-vitro effects of 4 of the principal components symbolizing the MD such as olive oil (polyphenol), red chili (capsaicin), tomato (lycopene) and red grapes (resveratrol) have been systematically reviewed. Several studies have demonstrated that polyphenols form olive oil, lycopene, resveratrol and capsaicin have multiple anticancer properties affecting several metabolic pathways involved in cancerogenesis, apoptosis, and metastasis in CRC cell lines. This review summarizes some of the most recent data potentially supportive of the use of MD in CRC chemoprevention, analyzing the in vitro effects of individual components of the MD on CRC cell development, progression, metastasis and apoptosis. Copyright © 2015 Elsevier Ltd. All rights reserved.

  1. Simultaneous analysis of 11 main active components in Cirsium setosum based on HPLC-ESI-MS/MS and combined with statistical methods.

    Science.gov (United States)

    Sun, Qian; Chang, Lu; Ren, Yanping; Cao, Liang; Sun, Yingguang; Du, Yingfeng; Shi, Xiaowei; Wang, Qiao; Zhang, Lantong

    2012-11-01

    A novel method based on high-performance liquid chromatography coupled with electrospray ionization tandem mass spectrometry was developed for simultaneous determination of the 11 major active components including ten flavonoids and one phenolic acid in Cirsium setosum. Separation was performed on a reversed-phase C(18) column with gradient elution of methanol and 0.1‰ acetic acid (v/v). The identification and quantification of the analytes were achieved on a hybrid quadrupole linear ion trap mass spectrometer. Multiple-reaction monitoring scanning was employed for quantification with switching electrospray ion source polarity between positive and negative modes in a single run. Full validation of the assay was carried out including linearity, precision, accuracy, stability, limits of detection and quantification. The results demonstrated that the method developed was reliable, rapid, and specific. The 25 batches of C. setosum samples from different sources were first determined using the developed method and the total contents of 11 analytes ranged from 1717.460 to 23028.258 μg/g. Among them, the content of linarin was highest, and its mean value was 7340.967 μg/g. Principal component analysis and hierarchical clustering analysis were performed to differentiate and classify the samples, which is helpful for comprehensive evaluation of the quality of C. setosum. © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  2. The Protective Effect of Lavender Essential Oil and Its Main Component Linalool against the Cognitive Deficits Induced by D-Galactose and Aluminum Trichloride in Mice

    Directory of Open Access Journals (Sweden)

    Pan Xu

    2017-01-01

    Full Text Available Lavender essential oil (LO is a traditional medicine used for the treatment of Alzheimer’s disease (AD. It was extracted from Lavandula angustifolia Mill. This study was designed to investigate the effects of lavender essential oil (LO and its active component, linalool (LI, against cognitive impairment induced by D-galactose (D-gal and AlCl3 in mice and to explore the related mechanisms. Our results revealed that LO (100 mg/kg or LI (100 mg/kg significantly protected the cognitive impairments as assessed by the Morris water maze test and step-though test. The mechanisms study demonstrated that LO and LI significantly protected the decreased activity of superoxide dismutase (SOD, glutathione peroxidase (GPX, and protected the increased activity of acetylcholinesterase (AChE and content of malondialdehyde (MDA. Besides, they protected the suppressed nuclear factor-erythroid 2-related factor 2 (Nrf2 and heme oxygenase-1 (HO-1 expression significantly. Moreover, the decreased expression of synapse plasticity-related proteins, calcium-calmodulin-dependent protein kinase II (CaMKII, p-CaMKII, brain-derived neurotrophic factor (BDNF, and TrkB in the hippocampus were increased with drug treatment. In conclusion, LO and its active component LI have protected the oxidative stress, activity of cholinergic function and expression of proteins of Nrf2/HO-1 pathway, and synaptic plasticity. It suggest that LO, especially LI, could be a potential agent for improving cognitive impairment in AD.

  3. Final environmental impact statement for the continued operation of the Pantex Plant and associated storage of nuclear weapon components. Volume 1 -- Main report

    International Nuclear Information System (INIS)

    1996-11-01

    This document assesses the potential environmental impacts over approximately 10 years of continued operation of Pantex Plant, including foreseeable projects and activities. For Pantex Plant, this document assesses the alternatives of No Action, Relocation of the storage of plutonium components (pits) resulting from nuclear weapon disassembly activities at Pantex Plant to another site, and the Proposed Action (Preferred Alternative) of continuing operations and increasing the quantity of pits in interim storage at Pantex Plant. For the Pit Storage Relocation Alternative, this document also assesses the potential environmental impacts to three DOE candidate sites and one Department of Defense candidate site that could be selected for the relocation of the nuclear component storage activities from Pantex Plant. Evaluations of site infrastructure, land resources, geology and soils, water resources, air quality, acoustics, biotic resources, cultural resources, socio-economic resources, intrasite transportation, waste management, human health, aircraft accidents, and environmental justice for Pantex Plant and the candidate sites are included in the assessment. The intersite transportation of nuclear and hazardous materials is also assessed

  4. Outline and status of ITER program

    International Nuclear Information System (INIS)

    Kishimoto, Hiroshi; Shimomura, Yasuo

    2002-01-01

    ITER is an international joint program for the next-step fusion experimental reactor which aims to demonstrate extended/steady-state fusion burn of deuterium-tritium plasmas and to demonstrate the fusion technologies in an integrated manner as well as to perform integrated testing of components required to utilize fusion energy for practical purposes. On the basis of the recent scientific and engineering achievements in the world-wide tokamak research, the Engineering Design Activities for nine years were fully completed in July 2001. The so-called compact ITER with a finite Q≥10 was proposed and its detailed engineering design was developed along the line of world fusion research. Large scale engineering research and development were completed for superconducting coils, remote-maintenance technology, etc.. The four ITER Parties (Japan, the European Union, the Soviet Federation, and Canada) have initiated the governmental negotiations for the joint implementation of ITER. (author)

  5. Assessment of radiation maps during activated divertor moving in the ITER building

    International Nuclear Information System (INIS)

    Ying Dongchuan; Zeng Qin; Qiu Yuefeng; Dang Tongqiang; Wu Yican; Loughlin, Michael

    2011-01-01

    As the main interface components between plasma and vacuum vessel, the divertor is foreseen to be removed to the hot cell for refurbishment during the 20 years of ITER operation. During this process, the activated divertor will cause a large increase of radiation in the ITER building. 3D analysis has been performed to assess the radiation maps throughout the ITER building for assisting the shielding design for personnel and sensitive equipment. The activation of the divertor has been determined by coupled neutron transport and inventory calculations, radiation maps have been obtained from gamma transport calculations. The neutron and gamma transport calculations have been performed by MCNP5 code with FENDL2.1library. The inventory calculations have been performed by FISPACT2007 code with EAF-2007 library. The results of these 3D decay gamma radiation maps are presented by pictures in this paper, including the biological dose maps and maps of heat deposition in electronic equipment.

  6. R and D of atmosphere detritiation system for ITER in JAEA

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Takumi, E-mail: hayashi.takumi@jaea.go.jp [Tritium Technol. Gr.: Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki Pref. 319-1195 (Japan); Iwai, Yasunori; Kobayashi, Kazuhiro; Nakamura, Hirofumi; Yamanishi, Toshihiko [Tritium Technol. Gr.: Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki Pref. 319-1195 (Japan); Perevezentsev, Alexander [Tritium Plant Gr.: ITER Organization, Cadarache Site (France)

    2010-12-15

    In order to establish an effective ITER atmosphere detritiation system (DS), JAEA (Japan Atomic Energy Agency) has investigated the performance and the durability of the system at various incident/accident conditions in support of finalizing the DS conceptual design through the ITER design review. The current DS as the Safety Important Component (SIC) has been discussed and mainly consists of catalytic reactors, wet scrubber columns (SCs), and blowers. The functional failure of the DS designed with SC was evaluated using a database of failure experiences of various valves, controllers, and components. Regarding the tritium release into the biggest confinement sector of the Tokamak gallery, this design is improved by more than two orders of magnitude compared to the original DS designed with molecular sieves (MSs) dryer beds in the 2001 design report. This improvement is achieved mainly by the reduction of frequency of valve operation, like MS dryers requiring periodical regeneration, and by the standardized module arrangement of the DS with SC.

  7. ITER towards the construction

    International Nuclear Information System (INIS)

    Shimomura, Y.

    2005-01-01

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

  8. Adjust of the residuals of the Arima model by means of the analysis of the residuals of the explanatory variables by means of the analysis of main components

    International Nuclear Information System (INIS)

    Bernal Suarez, Nestor Ricardo; Montealegre Bocanegra, Jose Edgar

    2000-01-01

    Based on the previous knowledge and understanding of the causality relationships between the fields of surface temperature of the Pacific and North and South tropical Atlantic oceans and rainfall behaviour in Colombia, we purport to model those relations with a (statistical) transfer model. This work is aimed at improving the adjustment of the model for the monthly mean rainfall registered in Funza (nearby the Capital Bogota). The residues of ARIMA models with six explanatory variables may contribute some percentage to the explanation of the total variability of rainfall as a consequence of their interrelationship. Such effect can be represented as a summary of the six variables, which can be achieved with principal components, taking into account that they are not mutually dependent, since they are white noise time series

  9. Power and particle control for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, S A; Cummings, J; Post, D E; Redi, M H [Princeton Univ., NJ (USA). Plasma Physics Lab.; Braams, B J [New York Univ., NY (USA). Courant Inst. of Mathematical Sciences; Brooks, J [Argonne National Lab., IL (USA); Engelmann, F; Pacher, G W; Pacher, H D [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.). NET Design Team; Harrison, M; Hotston, E [AEA Fusion, Culham (UK).

    1990-12-15

    Achievement of ITER's objectives, long-pulse ignited operation and nuclear component testing in quasi-steady-state, requires exhaust of power and helium ash, control of impurity content, and long lifetimes for plasma-facing components. In this paper we describe the data base and modeling results used to extrapolate present edge plasma parameters to ITER. Particular emphasis has been given to determining the uncertainties in predicted divertor performance. These analyses have been applied to four typical scenarios: A1 (ignited, reference Physics Phase), B1 (long pulse, hybrid, Technology Phase), B6 (steady-state, Technology Phase, impurity seeded) and B4 (steady-state, Technology Phase). 43 refs., 3 tabs.

  10. Study for reducing radioactive solid waste at ITER decommissioning period

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Shinichi; Araki, Masanori; Ohmori, Junji; Ohno, Isamu; Sato, Satoshi; Yamauchi, Michinori; Nishitani, Takeo [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-11-01

    It is one of the foremost goals for ITER to demonstrate the attractiveness with regard to safety and environmental potential. This implies that the radioactive materials and waste at decommissioning phase should carefully be treated with prescribed regulations. As possible activities during the Coordinated Technical Activity (CTA), the authors have performed a feasibility study for searching the possibility of effective reduction in the activated level as reasonably achievable as possible by taking account of minimum material changes while keeping original design concept and structure. Major induced activation in ITER comes from activated nickel and cobalt so that it is effective for the major structural components to minimize their material contents. Employing less Ni and Co steel in place of high-Ni austenitic stainless steel for blanket shield block, vacuum vessel shield material and TF coil casing has been considered as one of the effective plans to reduce the activated materials at the decommissioning phase. In this study, two less-Ni austenitic stainless steels are evaluated; one is high-Mn austenitic stainless steel JK2 which is developing for jacket material of ITER CS coil and the other is SS204L/ASTM-XM-11 which is also high-Mn steel specified in the popular standards such as American Society of Testing and Material (ASTM). Based on the material changes, activation analyses have been performed to investigate the possibility of reducing radioactive wastes. As a most impressive result, at 40 years after the termination some of main components such as a TF coil casing will reach to the clearance level which is specified by IAEA, and most components will be categorized into extremely low level waste except for limited components. These results will give the appropriate short decommissioning period that is assumed to start at 100 years after the termination in the original design. (author)

  11. ATHENA calculation model for the ITER-FEAT divertor cooling system. Final report with updates

    International Nuclear Information System (INIS)

    Eriksson, John; Sjoeberg, A.; Sponton, L.L.

    2001-05-01

    An ATHENA model of the ITER-FEAT divertor cooling system has been developed for the purpose of calculating and evaluating consequences of different thermal-hydraulic accidents as specified in the Accident Analysis Specifications for the ITER-FEAT Generic Site Safety Report. The model is able to assess situations for a variety of conceivable operational transients from small flow disturbances to more critical conditions such as total blackout caused by a loss of offsite and emergency power. The main objective for analyzing this type of scenarios is to determine margins against jeopardizing the integrity of the divertor cooling system components and pipings. The model of the divertor primary heat transport system encompasses the divertor cassettes, the port limiter systems, the pressurizer, the heat exchanger and all feed and return pipes of these components. The development was pursued according to practices and procedures outlined in the ATHENA code manuals using available modelling components such as volumes, junctions, heat structures and process controls

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  13. ITER status, design and material objectives

    International Nuclear Information System (INIS)

    Aymar, R.

    2002-01-01

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

  14. Tritium behavior in ITER beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-10-01

    The beryllium neutron multiplier in the ITER breeding blanket will generate tritium through transmutations. That tritium constitutes a safety hazard. Experiments evaluating tritium storage and release mechanisms have shown that most of the tritium comes out in a burst during thermal ramping. A small fraction of retained tritium is released by thermally activated processes. Analysis of recent experimental data shows that most of the tritium resides in helium bubbles. That tritium is released when the bubbles undergo swelling sufficient to develop porosity that connects with the surface. That appears to occur when swelling reaches about 10--15%. Other tritium appears to be stored chemically at oxide inclusions, probably as Be(OT) 2 . That component is released by thermal activation. There is considerable variation in published values for tritium diffusion through the beryllium and solubility in it. Data from experiments using highly irradiated beryllium from the Idaho National Engineering Laboratory showed diffusivity generally in line with the most commonly accepted values for fully dense material. Lower density material, planned for use in the ITER blanket may have very short diffusion times because of the open structure. The beryllium multiplier of the ITER breeding blanket was analyzed for tritium release characteristics using temperature and helium production figures at the midplane generated in support of the ITER Summer Workshop, 1990 in Garching. Ordinary operation, either in Physics or Technology phases, should not result in the release of tritium trapped in the helium bubbles. Temperature excursions above 600 degree C result in large-scale release of that tritium. 29 refs., 10 figs., 3 tabs

  15. Infrared laser diagnostics for ITER

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  16. 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

  17. Evaluation of the energy component in the main grid tariff and the use of marginal loss rates; Evaluering av energileddet i sentralnettstariffen og bruk av marginaltapssatser

    Energy Technology Data Exchange (ETDEWEB)

    Bjoerndal, Endre; Bjoerndal, Mette

    2010-11-15

    In this report, we make an evaluation of the energy component of the central grid tariff and the use of marginal loss rates. We will first review the theory of optimum transmission tariffs and describes how the marginal loss rates are calculated and how these are used in combination with the system price in the current transmission costs. We describe the trend in the Swedish marginal loss schedule, and we refer the very short development in the PJM area (U.S. east coast). As the Swedish and the Norwegian system is apparently based on the same principles, we problematize the lack of harmonization, since the systems actually are now very different. There is little doubt that Statnett has refined and improved methods for calculating the marginal loss rates, while ensuring that bottlenecks are still handled fairly easy by a few large price areas on Nord Pool Spot. We discuss the relationship between loss-tariff and congestion management, and we recommend that Statnett considering using the area price as the settlement price. This will make it much easier for operators to take account of losses. (Author)

  18. Capsular Polysaccharide is a Main Component of Mycoplasma ovipneumoniae in the Pathogen-Induced Toll-Like Receptor-Mediated Inflammatory Responses in Sheep Airway Epithelial Cells.

    Science.gov (United States)

    Jiang, Zhongjia; Song, Fuyang; Li, Yanan; Xue, Di; Deng, Guangcun; Li, Min; Liu, Xiaoming; Wang, Yujiong

    2017-01-01

    Mycoplasma ovipneumoniae ( M. ovipneumoniae ) is characterized as an etiological agent of primary atypical pneumonia that specifically infects sheep and goat. In an attempt to better understand the pathogen-host interaction between the invading M. ovipneumoniae and airway epithelial cells, we investigated the host inflammatory responses against capsular polysaccharide (designated as CPS) of M. ovipneumoniae using sheep bronchial epithelial cells cultured in an air-liquid interface (ALI) model. Results showed that CPS derived from M. ovipneumoniae could activate toll-like receptor- (TLR-) mediated inflammatory responses, along with an elevated expression of nuclear factor kappa B (NF- κ B), activator protein-1 (AP-1), and interferon regulatory factor 3 (IRF3) as well as various inflammatory-associated mediators, representatively including proinflammatory cytokines, such as IL1 β , TNF α , and IL8, and anti-inflammatory cytokines such as IL10 and TGF β of TLR signaling cascade. Mechanistically, the CPS-induced inflammation was TLR initiated and was mediated by activations of both MyD88-dependent and MyD88-independent signaling pathways. Of importance, a blockage of CPS with specific antibody led a significant reduction of M. ovipneumoniae -induced inflammatory responses in sheep bronchial epithelial cells. These results suggested that CPS is a key virulent component of M. ovipneumoniae , which may play a crucial role in the inflammatory response induced by M. ovipneumoniae infections.

  19. Capsular Polysaccharide is a Main Component of Mycoplasma ovipneumoniae in the Pathogen-Induced Toll-Like Receptor-Mediated Inflammatory Responses in Sheep Airway Epithelial Cells

    Directory of Open Access Journals (Sweden)

    Zhongjia Jiang

    2017-01-01

    Full Text Available Mycoplasma ovipneumoniae (M. ovipneumoniae is characterized as an etiological agent of primary atypical pneumonia that specifically infects sheep and goat. In an attempt to better understand the pathogen-host interaction between the invading M. ovipneumoniae and airway epithelial cells, we investigated the host inflammatory responses against capsular polysaccharide (designated as CPS of M. ovipneumoniae using sheep bronchial epithelial cells cultured in an air-liquid interface (ALI model. Results showed that CPS derived from M. ovipneumoniae could activate toll-like receptor- (TLR- mediated inflammatory responses, along with an elevated expression of nuclear factor kappa B (NF-κB, activator protein-1 (AP-1, and interferon regulatory factor 3 (IRF3 as well as various inflammatory-associated mediators, representatively including proinflammatory cytokines, such as IL1β, TNFα, and IL8, and anti-inflammatory cytokines such as IL10 and TGFβ of TLR signaling cascade. Mechanistically, the CPS-induced inflammation was TLR initiated and was mediated by activations of both MyD88-dependent and MyD88-independent signaling pathways. Of importance, a blockage of CPS with specific antibody led a significant reduction of M. ovipneumoniae-induced inflammatory responses in sheep bronchial epithelial cells. These results suggested that CPS is a key virulent component of M. ovipneumoniae, which may play a crucial role in the inflammatory response induced by M. ovipneumoniae infections.

  20. Perl Modules for Constructing Iterators

    Science.gov (United States)

    Tilmes, Curt

    2009-01-01

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

  1. ITER definition phase

    International Nuclear Information System (INIS)

    1989-01-01

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

  2. Antibacterial activity and mechanism of action of Monarda punctata essential oil and its main components against common bacterial pathogens in respiratory tract.

    Science.gov (United States)

    Li, Hong; Yang, Tian; Li, Fei-Yan; Yao, Yan; Sun, Zhong-Min

    2014-01-01

    The aim of the current research work was to study the chemical composition of the essential oil of Monarda punctata along with evaluating the essential oil and its major components for their antibacterial effects against some frequently encountered respiratory infection causing pathogens. Gas chromatographic mass spectrometric analysis revealed the presence of 13 chemical constituents with thymol (75.2%), p-cymene (6.7%), limonene (5.4), and carvacrol (3.5%) as the major constituents. The oil composition was dominated by the oxygenated monoterpenes. Antibacterial activity of the essential oil and its major constituents (thymol, p-cymene, limonene) was evaluated against Streptococcus pyogenes, methicillin-resistant Staphylococcus aureus (MRSA), Streptococcus pneumoniae, Haemophilus influenzae and Escherichia coli. The study revealed that the essential oil and its constituents exhibited a broad spectrum and variable degree of antibacterial activity against different strains. Among the tested strains, Streptococcus pyogenes, Escherichia coli and Streptococcus pneumoniae were the most susceptible bacterial strain showing lowest MIC and MBC values. Methicillin-resistant Staphylococcus aureus was the most resistant bacterial strain to the essential oil treatment showing relatively higher MIC and MBC values. Scanning electron microscopy revealed that the essential oil induced potent and dose-dependent membrane damage in S. pyogenes and MRSA bacterial strains. The reactive oxygen species generated by the Monarda punctata essential oil were identified using 2', 7'-dichlorofluorescein diacetate (DCFDA).This study indicated that the Monarda punctata essential oil to a great extent and thymol to a lower extent triggered a substantial increase in the ROS levels in S. pyogenes bacterial cultures which ultimately cause membrane damage as revealed by SEM results.

  3. Sodium metabisulphite, a preservative agent, decreases the heart capillary volume and length, and curcumin, the main component of Curcuma longa, cannot protect it.

    Science.gov (United States)

    Noorafshan, A; Asadi-Golshan, R; Monjezi, S; Karbalay-Doust, S

    2014-01-01

    Sodium metabisulphite is used as an antioxidant agent in many pharmaceutical formulations. It is extensively used as a food preservative and disinfectant. It has been demonstrated that sulphite exposure can affect some organs. Curcumin, the main element of Curcuma longa, has been identified to have multiple protective properties. The present study extends the earlier works to quantitative evaluation of the effects of sulphite and curcumin on the heart structure using stereological methods. In this study, 28 rats were randomly divided into four experimental groups. The rats in groups I to IV received distilled water (group I), sodium metabisulphite (25 mg/ kg/day) (group II), curcumin (100 mg/kg/day) (group III), and sodium metabisulphite+curcumin (group IV), respectively, for 8 weeks. The left ventricle was subjected to stereological methods to estimate the quantitative parameters of the myocardium. A 20 % decrease was observed in the total volume of ventricular tissue in the sulphite-treated animals compared to the distilled water treatment (P curcumin did not protect the animals against the structural changes of the ventricle. Sulphite, as a preservative food agent, reduced the length and volume of the ventricular capillaries and curcumin could not protect them.

  4. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.

    1995-01-01

    Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

  5. ITER EDA and technology

    International Nuclear Information System (INIS)

    Baker, C.C.

    2001-01-01

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

  6. Main findings

    International Nuclear Information System (INIS)

    2014-01-01

    Licensing regimes vary from country to country. When the license regime involves several regulators and several licenses, this may lead to complex situations. Identifying a leading organisation in charge of overall coordination including preparation of the licensing decision is a useful practice. Also, if a stepwise licensing process is implemented, it is important to fix in legislation decisions and/or time points and to identify the relevant actors. There is considerable experience in civil and mining engineering that can be applied when constructing a deep geological disposal facility. Specific challenges are, however, the minimization of disturbances to the host rock and the understanding of its long-term behavior. Construction activities may affect the geo-hydraulic and geochemical properties of the various system components which are important safety features of the repository system. Clearly defined technical specifications and an effective quality management plan are important in ensuring successful repository implementation which is consistent with safety requirements. Monitoring plan should also be defined in advance. The regulatory organization should prepare itself to the licensing review before construction by allocating sufficient resources. It should increase its competence, e.g., by interacting early with the implementer and through its own R and D. This will allow the regulator to define appropriate technical conditions associated to the construction license and to elaborate a relevant inspection plan of the construction work. After construction, obtaining the operational license is the most important and crucial step. Main challenges include (a) establishing sufficient confidence so that the methods for closing the individual disposal units comply with the safety objectives and (b) addressing the issue of ageing of materials during a 50-100 years operational period. This latter challenge is amplified when reversibility/retrievability is required

  7. 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.

  8. Toward construction of ITER

    International Nuclear Information System (INIS)

    Shimomura, Yasuo

    2005-01-01

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

  9. ITER Status and Plans

    Science.gov (United States)

    Greenfield, Charles M.

    2017-10-01

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

  10. Matlab modeling of ITER CODAC

    International Nuclear Information System (INIS)

    Pangione, L.; Lister, J.B.

    2008-01-01

    The ITER CODAC (COntrol, Data Access and Communication) conceptual design resulted from 2 years of activity. One result was a proposed functional partitioning of CODAC into different CODAC Systems, each of them partitioned into other CODAC Systems. Considering the large size of this project, simple use of human language assisted by figures would certainly be ineffective in creating an unambiguous description of all interactions and all relations between these Systems. Moreover, the underlying design is resident in the mind of the designers, who must consider all possible situations that could happen to each system. There is therefore a need to model the whole of CODAC with a clear and preferably graphical method, which allows the designers to verify the correctness and the consistency of their project. The aim of this paper is to describe the work started on ITER CODAC modeling using Matlab/Simulink. The main feature of this tool is the possibility of having a simple, graphical, intuitive representation of a complex system and ultimately to run a numerical simulation of it. Using Matlab/Simulink, each CODAC System was represented in a graphical and intuitive form with its relations and interactions through the definition of a small number of simple rules. In a Simulink diagram, each system was represented as a 'black box', both containing, and connected to, a number of other systems. In this way it is possible to move vertically between systems on different levels, to show the relation of membership, or horizontally to analyse the information exchange between systems at the same level. This process can be iterated, starting from a global diagram, in which only CODAC appears with the Plant Systems and the external sites, and going deeper down to the mathematical model of each CODAC system. The Matlab/Simulink features for simulating the whole top diagram encourage us to develop the idea of completing the functionalities of all systems in order to finally have a full

  11. Toxicity of the main electronic cigarette components, propylene glycol, glycerin, and nicotine, in Sprague-Dawley rats in a 90-day OECD inhalation study complemented by molecular endpoints.

    Science.gov (United States)

    Phillips, Blaine; Titz, Bjoern; Kogel, Ulrike; Sharma, Danilal; Leroy, Patrice; Xiang, Yang; Vuillaume, Grégory; Lebrun, Stefan; Sciuscio, Davide; Ho, Jenny; Nury, Catherine; Guedj, Emmanuel; Elamin, Ashraf; Esposito, Marco; Krishnan, Subash; Schlage, Walter K; Veljkovic, Emilija; Ivanov, Nikolai V; Martin, Florian; Peitsch, Manuel C; Hoeng, Julia; Vanscheeuwijck, Patrick

    2017-11-01

    While the toxicity of the main constituents of electronic cigarette (ECIG) liquids, nicotine, propylene glycol (PG), and vegetable glycerin (VG), has been assessed individually in separate studies, limited data on the inhalation toxicity of them is available when in mixtures. In this 90-day subchronic inhalation study, Sprague-Dawley rats were nose-only exposed to filtered air, nebulized vehicle (saline), or three concentrations of PG/VG mixtures, with and without nicotine. Standard toxicological endpoints were complemented by molecular analyses using transcriptomics, proteomics, and lipidomics. Compared with vehicle exposure, the PG/VG aerosols showed only very limited biological effects with no signs of toxicity. Addition of nicotine to the PG/VG aerosols resulted in effects in line with nicotine effects observed in previous studies, including up-regulation of xenobiotic enzymes (Cyp1a1/Fmo3) in the lung and metabolic effects, such as reduced serum lipid concentrations and expression changes of hepatic metabolic enzymes. No toxicologically relevant effects of PG/VG aerosols (up to 1.520  mg PG/L + 1.890 mg VG/L) were observed, and no adverse effects for PG/VG/nicotine were observed up to 438/544/6.6 mg/kg/day. This study demonstrates how complementary systems toxicology analyses can reveal, even in the absence of observable adverse effects, subtoxic and adaptive responses to pharmacologically active compounds such as nicotine. Copyright © 2017 The Authors. Published by Elsevier Ltd.. All rights reserved.

  12. Ceramics radiation effects issues for ITER

    International Nuclear Information System (INIS)

    Zinkle, S.J.

    1993-01-01

    The key radiation effects issues associated with the successful operation of ceramic materials in components of the planned International Thermonuclear Experimental Reactor (ITER) are discussed. Radiation-induced volume changes and degradation of the mechanical properties should not be a serious issue for the fluences planned for ITER. On the other hand, radiation-induced electrical degradation effects may severely limit the allowable exposure of ceramic insulators. Degradation of the loss tangent and thermal conductivity may also restrict the location of some components such as ICRH feedthrough insulators to positions far away from the first wall. In-situ measurements suggest that the degradation of physical properties in ceramics during irradiation is greater than that measured in postirradiation tests. Additional in-situ data during neutron irradiation are needed before engineering designs for ITER can be finalized

  13. 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.

  14. Selection of plasma facing materials for ITER

    International Nuclear Information System (INIS)

    Ulrickson, M.; Barabash, V.; Chiocchio, S.

    1996-01-01

    ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m 2 for normal operation with transients to 15 MW/m 2 for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final section of plasma facing materials for the ITER divertor will not be made until the end of the EDA

  15. Conceptual design of a test facility for the remote handling operations of the ITER Test Blanker Modules

    International Nuclear Information System (INIS)

    Marqueta, A.; Garcia, I.; Gomez, A.; Garcia, L.; Sedano, E.; Fernandez, I.

    2012-01-01

    Conceptual Design of a test facility for the remote handling operations of the ITER Test Blanket Modules. Conditions inside a fusion reactor are incompatible with conventional manual maintenance tasks. the same applies for ancillary equipment. As a consequence, it will become necessary to turn to remote visualization and remote handling techniques, which will have in consideration the extreme conditions, both physical and operating, of ITER. Main goal of the project has been the realization of the conceptual design for the test facility for the Test Blanket Modules of ITER and their associated systems, related to the Remote Handling operations regarding the Port Cell area. Besides the definition of the operations and the specification of the main components and ancillary systems of the TBM graphical simulation have been used for the design, verification and validation of the remote handling operations. (Author)

  16. ITER PF6 double pancakes winding line

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

  17. ITER CTA newsletter. No. 6

    International Nuclear Information System (INIS)

    2002-01-01

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

  18. ITER CTA newsletter. No. 2

    International Nuclear Information System (INIS)

    2001-10-01

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

  19. 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.

  20. 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.

  1. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    2001-01-01

    The divertor ''Large Project'' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  2. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    1999-01-01

    The divertor 'Large Project' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  3. ITER management advisory committee (MAC) meeting in Naka

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    2000-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held on 28 June 2000 in Moskow, Russia. The main topics were the consideration of the report by the director on the ITER EDA status, the review of the work program, the review of the joint fund, the review of a schedule of ITER meetings and initial discussion and consideration on the disposition of R and D hardware and facilities and other dispositions relating to the termination of the EDA

  4. The ITER management advisory committee (MAC) meeting in Cadarache

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    1999-01-01

    The ITER management advisory committee meeting was held on 8 March in Cadarache, France. The main topics were the ITER EDA Status Report in the period between the ITER Meeting in Yokohama (October 1990) and February 1999. In particular, the management advisory committee shares the director's concern about the uncertainties against which the whole project operates at present. They also noted that the definition of an appropriate framework for continued US involvement in on-going projects is a very important issue

  5. 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

  6. ITER-FEAT operation

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  7. Qualification Test for Korean Mockups of ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Kim, S. K.; Lee, D. W.; Bae, Y. D.; Hong, B. G.; Jung, H. K.; Jung, Y. I.; Park, J. Y.; Jeong, Y. H.; Choi, B. K.; Kim, B. Y.

    2009-01-01

    ITER First Wall (FW) includes the beryllium armor tiles joined to CuCrZr heat sink with stainless steel cooling tubes. This first wall panels are one of the critical components in the ITER machine with the surface heat flux of 0.5 MW/m 2 or above. So qualification program needs to be performed with the goal to qualify the joining technologies required for the ITER First Wall. Based on the results of tests, the acceptance of the developed joining technologies will be established. The results of this qualification test will affect the final selection of the manufacturers for the ITER First Wall

  8. 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

  9. Comparative analysis of main bio-active components in the herb pair Danshen-Honghua and its single herbs by ultra-high performance liquid chromatography coupled to triple quadrupole tandem mass spectrometry.

    Science.gov (United States)

    Qu, Cheng; Pu, Zong-Jin; Zhou, Gui-Sheng; Wang, Jun; Zhu, Zhen-Hua; Yue, Shi-Jun; Li, Jian-Ping; Shang, Li-Li; Tang, Yu-Ping; Shi, Xu-Qin; Liu, Pei; Guo, Jian-Ming; Sun, Jing; Tang, Zhi-Shu; Zhao, Jing; Zhao, Bu-Chang; Duan, Jin-Ao

    2017-09-01

    A sensitive, reliable, and powerful ultra-high performance liquid chromatography coupled to triple quadrupole tandem mass spectrometry method was developed for simultaneous quantification of the 15 main bio-active components including phenolic acids and flavonoids within 13 min for the first time. The proposed method was first reported and validated by good linearity (r 2  > 0.9975), limit of detection (1.12-7.01 ng/mL), limit of quantification (3.73-23.37 ng/mL), intra- and inter-day precisions (RSD ≤ 1.92%, RSD ≤ 2.45%), stability (RSD ≤ 5.63%), repeatability (RSD ≤ 4.34%), recovery (96.84-102.12%), and matrix effects (0.92-1.02). The established analytical methodology was successfully applied to comparative analysis of main bio-active components in the herb pair Danshen-Honghua and its single herbs. Compared to the single herb, the content of most flavonoid glycosides was remarkably increased in their herb pair, and main phenolic acids were decreased, conversely. The content changes of the main components in the herb pair supported the synergistic effects on promoting blood circulation and removing blood stasis. The results provide a scientific basis and reference for the quality control of Danshen-Honghua herb pair and the drug interactions based on variation of bio-active components in herb pairs. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  10. 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

  11. Significance of ITER IWS Material Selection and Qualification

    Science.gov (United States)

    Mehta, Bhoomi K.; Raval, Jigar; Maheshwari, Abha; Laad, Rahul; Singh, Gurlovleen; Pathak, Haresh

    2017-04-01

    In-Wall Shielding (IWS) is one of the important components of ITER Vacuum Vessel (VV) which fills the space between double walls of VV with cooling water. Procurement Arrangement (PA) for IWS has been signed with Indian Domestic Agency (INDA). Procurement of IWS materials, fabrication of IWS blocks and its delivery to respective Domestic Agency (DA) and ITER Organization (IO) are the main scope of this PA. Hence, INDIA is the only country which is contributing to VV IWS among all seven ITER partners. The main functions of the IWS are to provide Neutron Shielding with blanket, VV shells and water during plasma operations and to reduce ripple of the Toroidal Magnetic Field. To meet these functional requirements IWS blocks are made up of special materials (Borated Steels SS304 B4 & SS304 B7, Ferritic Steels SS 430, Austenitic Steel SS 316 L (N)-IG, XM-19 and Inconel-625) which are qualified, reliable and traceable for the design assessment. The choice of these materials has a significant influence on performance, maintainability, licensing, detailed design parameters and waste disposal. The main reasons for the materials selected for IWS are its high mechanical strength at operating temperatures, water chemistry properties, excellent fabrication characteristics and low cost relative to other similar materials. All the materials are qualified with respect to their respective codes (ASTM/EN standards with additional requirements as described in RCC-MR code 2007) and ITER requirements. Agreed Notified Body (ANB) has control conformity of materials certificates with approved material specification and traceability procedure for Safety Important Component (SIC). The procurement strategy for all the IWS materials has been developed in close collaboration with IO, ANB and Industries as per Product Procurement Specification (PPS). The R&D for sample, bulk material production, testing, inspection and handling as required are carried out by IN DA and IO. At present almost all

  12. 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.

  13. The ITER bolometer diagnostic: Status and plans

    International Nuclear Information System (INIS)

    Meister, H.; Giannone, L.; Horton, L. D.; Raupp, G.; Zeidner, W.; Grunda, G.; Kalvin, S.; Fischer, U.; Serikov, A.; Stickel, S.; Reichle, R.

    2008-01-01

    A consortium consisting of four EURATOM Associations has been set up to develop the project plan for the full development of the ITER bolometer diagnostic and to continue urgent R and D activities. An overview of the current status is given, including detector development, line-of-sight optimization, performance analysis as well as the design of the diagnostic components and their integration in ITER. This is complemented by the presentation of plans for future activities required to successfully implement the bolometer diagnostic, ranging from the detector development over diagnostic design and prototype testing to RH tools for calibration.

  14. The ITER bolometer diagnostic: Status and plansa)

    Science.gov (United States)

    Meister, H.; Giannone, L.; Horton, L. D.; Raupp, G.; Zeidner, W.; Grunda, G.; Kalvin, S.; Fischer, U.; Serikov, A.; Stickel, S.; Reichle, R.

    2008-10-01

    A consortium consisting of four EURATOM Associations has been set up to develop the project plan for the full development of the ITER bolometer diagnostic and to continue urgent R&D activities. An overview of the current status is given, including detector development, line-of-sight optimization, performance analysis as well as the design of the diagnostic components and their integration in ITER. This is complemented by the presentation of plans for future activities required to successfully implement the bolometer diagnostic, ranging from the detector development over diagnostic design and prototype testing to RH tools for calibration.

  15. ITER central solenoid manufacturing R and D

    International Nuclear Information System (INIS)

    Jay Jayakumar, R.; Tsuji, H.; Ohsaki, O.

    2001-01-01

    The International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA) includes the development of high performance superconductors, high current joints between superconducting cables and insulating materials. Also in the EDA, the resulting products of this R and D are incorporated in a Central Solenoid Model Coil which utilizes full size conductors. The manufacturing of the model coil and components has led to the development of the design, materials, tooling and process which are fully applicable to the manufacture of the ITER relevant CS coil. The R and D is essentially complete and final stages of the CS Model Coil manufacturing are underway. (author)

  16. ITER central solenoid manufacturing R and D

    International Nuclear Information System (INIS)

    Jayakumar, R.J.; Tsuji, H.; Ohsaki, O.

    1999-01-01

    The International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA) includes the development of high performance superconductors, high current joints between superconducting cables and insulating materials. Also in the EDA, the resulting products of this R and D are incorporated in a Central Solenoid Model Coil which utilizes full size conductors. The manufacturing of the model coil and components has led to the development of the design, materials, tooling and process which are fully applicable to the manufacture of the ITER relevant CS coil. The R and D is essentially complete and final stages of the CS Model Coil manufacturing are underway. (author)

  17. 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.

  18. Progress of the ITER Correction Coils in China

    CERN Document Server

    Wei, J; Han, S; Yu, X; Du, S; Li, C; Fang, C; Wang, L; Zheng, W; Liu, L; Wen, J; Li, H; Libeyre, P; Dolgetta, N; Cormany, C; Sgobba, S

    2014-01-01

    The ITER Correction Coils (CC) include three sets of six coils each, distributed symmetrically around the tokamak to correct error fields. Each pair of coils, located on opposite sides of the tokamak, is series connected with polarity to produce asymmetric fields. The manufacturing of these superconducting coils is undergoing qualification of the main fabrication processes: winding into multiple pancakes, welding helium inlet/outlet on the conductor jacket, turn and ground insulation, vacuum pressure impregnation, inserting into an austenitic stainless steel case, enclosure welding, and assembling the terminal service box. It has been proceeding by an intense phase of R\\&D, trials tests, and final adjustment of the tooling. This paper mainly describes the progress in ASIPP for the CC manufacturing process before and on qualification phase and the status of corresponding equipment which are ordered or designed for each process. Some test results for the key component and procedure are also presented.

  19. Earthly sun called ITER

    International Nuclear Information System (INIS)

    Pozdeyev, Mikhail

    2002-01-01

    Full text: Participating in the film are Academicians Velikhov and Glukhikh, Mr. Filatof, ITER Director from Russia, Mr. Sannikov from Kurchatov Institute. The film tells about the starting point of the project (Mr. Lavrentyev), the pioneers of the project (Academicians Tamme, Sakharov, Artsimovich) and about the situation the project is standing now. Participating in [ITER now are the US, Russia, Japan and the European Union. There are two associated members as well - Kazakhstan and Canada. By now the engineering design phase has been finished. Computer animation used in the video gives us the idea how the first thermonuclear reactor based on famous Russian TOKOMAK works. (author)

  20. 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

  1. Iterated multidimensional wave conversion

    International Nuclear Information System (INIS)

    Brizard, A. J.; Tracy, E. R.; Johnston, D.; Kaufman, A. N.; Richardson, A. S.; Zobin, N.

    2011-01-01

    Mode conversion can occur repeatedly in a two-dimensional cavity (e.g., the poloidal cross section of an axisymmetric tokamak). We report on two novel concepts that allow for a complete and global visualization of the ray evolution under iterated conversions. First, iterated conversion is discussed in terms of ray-induced maps from the two-dimensional conversion surface to itself (which can be visualized in terms of three-dimensional rooms). Second, the two-dimensional conversion surface is shown to possess a symplectic structure derived from Dirac constraints associated with the two dispersion surfaces of the interacting waves.

  2. Physics fundamentals for ITER

    International Nuclear Information System (INIS)

    Rosenbluth, M.N.

    1999-01-01

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

  3. Design of coolant distribution system (CDS) for ITER PF AC/DC converter

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Bin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Zhiquan, E-mail: zhquansong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Fu, Peng; Xu, Xuesong; Li, Chuan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Min; Dong, Lin [China International Nuclear Fusion Energy Program Execution Center, Beijing 100862 (China)

    2016-10-15

    Highlights: • System process and arrangement has been proposed to meet the multiple requirements from the converter system. • Thermal hydraulic analysis model has been developed to size and predict the system operation behavior. • Prototype test has been performed to validate the proposed design methodology. - Abstract: The Poloidal Field (PF) converter unit, playing an essential role in the plasma shape and position control in vertical and horizontal direction, which is an important part of ITER power supply system. As an important subsystem of the converter unit, the coolant distribution system has the function to distribute the cooling water from ITER component cooling water system (CCWS) to its main components at the required flow rate, pressure and temperature. This paper presents the thermal hydraulic design of coolant distribution system for the ITER PF converter unit. Different operational requirements of the PF converter unit regarding flow rate, temperature and pressure have been analyzed to design the system process and arrangement. A thermal-hydraulic analysis model has been built to size the system and predict the flow rate and temperature distribution of the system under the normal operation. Based on the system thermal-hydraulic analysis results, the system pressure profile has been plotted to evaluate the pressure behavior along each client flow path. A CDS prototype for the ITER PF converter has been constructed and some experiments have been performed on it. A good agreement of the flow distribution and temperature behavior between the simulated and test results validate the proposed design methodology.

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

    International Nuclear Information System (INIS)

    Mitchell, N.

    2006-01-01

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

  5. Study of the layout plan in the tokamak complex building for ITER

    International Nuclear Information System (INIS)

    Sato, Kazuyoshi; Yagenji, Akira; Sekiya, Shigeki; Takahashi, Hideo; Tamura, Kousaku; Neyatani, Yuzuru; Hashimoto, Masayoshi; Ogino, Shunji; Nagamatsu, Nobuhide; Motohashi, Keiichi; Uehara, Masaharu; Kataoka, Takahiro; Ohashi, Hironori

    2006-03-01

    This report summarizes study of the layout plan in the ITER Tokamak complex building as an invite to set up its plant in Japan. To draw up this arrangement plan, final design report (FDR), which was designed for main components and determined for the non-site specific design, was reconstructed systematically for the Japanese site. A supplementary design was performed for the insufficiency parts of FDR. An additional study was also performed for the adaptation of a regulatory framework including technical safety requirements in Japan. We proposed the tokamak complex building with seismic isolation to combine with the hot cell building. Through the studies, a layout plan has been constructed including maintenance plan for personnel access and component route with in the building from assembly to operation period. This layout plan would be used as a basis in the construction period, although final decision will be done by ITER organization. (author)

  6. Integration of an iterative methodology for exergoeconomic improvement of thermal systems with a process simulator

    International Nuclear Information System (INIS)

    Vieira, Leonardo S.; Donatelli, Joao L.; Cruz, Manuel E.

    2004-01-01

    In this paper, we present the development and automated implementation of an iterative methodology for exergoeconomic improvement of thermal systems integrated with a process simulator, so as to be applicable to real, complex plants. The methodology combines recent available exergoeconomic techniques with new qualitative and quantitative criteria for the following tasks: (i) identification of decision variables that affect system total cost and exergetic efficiency; (ii) hierarchical classification of components; (iii) identification of predominant terms in the component total cost; and (iv) choice of main decision variables in the iterative process. To show the strengths and potential advantages of the proposed methodology, it is here applied to the benchmark CGAM cogeneration system. The results obtained are presented and discussed in detail and are compared to those reached using a mathematical optimization procedure

  7. Physics research needs for ITER

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1995-01-01

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

  8. Progress in licensing ITER in Cadarache

    International Nuclear Information System (INIS)

    Rodriguez-Rodrigo, Lina; Girard, Jean-Philippe; Uzan-Elbez, Joelle; Marbach, Gabriel; Garin, Pascal; Rosanvallon, Sandrine

    2005-01-01

    The licensing procedure for ITER in Europe in the framework of the French regulations is a non-prescriptive approach based on a continuous dialogue between the nuclear installation owner (or its representative) and the safety authority. In this paper, the licensing procedure and main safety issues, which are being studied in this process, are presented

  9. On varitional iteration method for fractional calculus

    Directory of Open Access Journals (Sweden)

    Wu Hai-Gen

    2017-01-01

    Full Text Available Modification of the Das’ variational iteration method for fractional differential equations is discussed, and its main shortcoming involved in the solution process is pointed out and overcome by using fractional power series. The suggested computational procedure is simple and reliable for fractional calculus.

  10. Twelfth meeting of the ITER physics expert group on diagnostics

    International Nuclear Information System (INIS)

    Costley, A.E.; Donne, A.J.H.

    2000-01-01

    The main technical objectives of the meeting were to review the present status of ITER and to determine any required changes in the specifications for plasma measurements; to review the progress and develop plans for meeting the goals of the voluntary R and D tasks approved by the ITER Physics Committee within the Parties; to review and plan the work of the five specialists electronic working groups, and to hear reports of ITER relevant diagnostic developments in the Party Laboratories and assess their possible application to ITER

  11. Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

    Directory of Open Access Journals (Sweden)

    G. De Temmerman

    2017-08-01

    Full Text Available As a licensed nuclear facility, ITER must limit the in-vessel tritium (T retention to reduce the risks of potential release during accidents, the inventory limit being set at 1kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513K for the FW and 623K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.

  12. Main Memory

    NARCIS (Netherlands)

    P.A. Boncz (Peter); L. Liu (Lei); M. Tamer Özsu

    2008-01-01

    htmlabstractPrimary storage, presently known as main memory, is the largest memory directly accessible to the CPU in the prevalent Von Neumann model and stores both data and instructions (program code). The CPU continuously reads instructions stored there and executes them. It is also called Random

  13. Divertor cassette movers prototypes for ITER

    International Nuclear Information System (INIS)

    Bogusch, E.; Batz, R.; Bieber, O.; Gottfried, R.; Cerdan, G.

    1998-01-01

    Following competitive tendering, in October 1996 Siemens was contracted by the European Commission to design and supply an assembly of four Divertor Cassette Movers Prototypes including the control and command systems for the movers proper. The assembly consisting of one Cassette Toroidal Mover (CTM), one Radial Mover Tractor (TRC), one Second Cassette Carrier (SCC), and one Radial Cassette Carrier (RCC) represents key components of the Divertor Test Platform at Brasimone, one of the seven large R+D projects for ITER. By detailed design, high-precision manufacturing and testing of these devices, Siemens contributed to the verification of an important task within the European R and D program towards ITER construction. Replacement of the divertor cassettes is a scheduled maintenance operation throughout the life of ITER. The successful fabrication and testing of the Divertor Cassette Movers Prototypes is all important milestone to verify this delicate operation. (authors)

  14. Structural analysis for the joint of the ITER ELM coil

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

  15. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    Science.gov (United States)

    Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2009-04-01

    The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

  16. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)], E-mail: bazylev@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe (Germany); Landman, I.; Pestchanyi, S. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Loarte, A. [ITER Organisation, Cadarache, 13108 Saint Paul Lez Durance Cedex (France)

    2009-04-30

    The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

  17. Erosion simulation of first wall beryllium armour after ITER transient heat loads and runaway electrons action

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B., E-mail: boris.bazylev@kit.edu [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Igitkhanov, Yu.; Landman, I.; Pestchanyi, S. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Loarte, A. [ITER Organisation, Cadarache, 13108 Saint Paul Lez Durance Cedex (France)

    2011-10-01

    Beryllium is foreseen as plasma facing armour for the first wall (FW) in ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) and runaway electrons impact are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J x B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated.

  18. Erosion simulation of first wall beryllium armour after ITER transient heat loads and runaway electrons action

    International Nuclear Information System (INIS)

    Bazylev, B.; Igitkhanov, Yu.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2011-01-01

    Beryllium is foreseen as plasma facing armour for the first wall (FW) in ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) and runaway electrons impact are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J x B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated.

  19. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    International Nuclear Information System (INIS)

    Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2009-01-01

    The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

  20. Iterative List Decoding

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom; Hjaltason, Johan

    2005-01-01

    We analyze the relation between iterative decoding and the extended parity check matrix. By considering a modified version of bit flipping, which produces a list of decoded words, we derive several relations between decodable error patterns and the parameters of the code. By developing a tree...... of codewords at minimal distance from the received vector, we also obtain new information about the code....