WorldWideScience

Sample records for iter blanket maintenance

  1. 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.

  2. 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

  3. Study on compact design of remote handling equipment for ITER blanket maintenance

    International Nuclear Information System (INIS)

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

    2006-03-01

    In the ITER, the neutrons created by D-T reactions activate structural materials, and thereby, the circumstance in the vacuum vessel is under intense gamma radiation field. Thus, the in-vessel components such as blanket are handled and replaced by remote handling equipment. The objective of this report is to study the compactness of the remote handling equipment (a vehicle/manipulator) for the ITER blanket maintenance. In order to avoid the interferences between the blanket and the equipment during blanket replacement in the restricted vacuum vessel, a compact design of the equipment is required. Therefore, the compact design is performed, including kinematic analyses aiming at the reduction of the sizes of the vehicle equipped with a manipulator handling the blanket and the rail for the vehicle traveling in the vacuum vessel. Major results are as follows: 1. The compact vehicle/manipulator is designed concentration on the reduction of the rail size and simplification of the guide roller mechanism as well as the reduction of the gear diameter for vehicle rotation around the rail. Height of the rail is reduced from 500 mm to 400 mm by a parameter survey for weight, stiffness and stress of the rail. The roller mechanism is divided into two simple functional mechanisms composed of rollers and a pad, that is, the rollers support relatively light loads during rail deployment and vehicle traveling while a pad supports heavy loads during blanket replacement. Regarding the rotation mechanism, the double helical gear is adopted, because it has higher contact ratio than the normal spur gear and consequently can transfer higher force. The smaller double helical gear, 996 mm in diameter, can achieve 26% higher output torque, 123.5 kN·m, than that of the original spur gear of 1,460 mm in diameter, 98 kN·m. As a result, the manipulator becomes about 30% lighter, 8 tons, than the original weight, 11.2 tons. 2. Based on the compact design of the vehicle/manipulator, the

  4. Tritium and heat management in ITER Test Blanket Systems port cell for maintenance operations

    Energy Technology Data Exchange (ETDEWEB)

    Giancarli, L.M., E-mail: luciano.giancarli@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Iseli, M.; Lepetit, L.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Livingston, D. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom); Nevière, J.C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Pascal, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ricapito, I. [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Wyse, S. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom)

    2014-10-15

    Highlights: •The ITER TBM Program is one of the ITER missions. •We model a TBM port cell with CFD to optimize the design choices. •The heat and tritium releases management in TBM port cells has been optimized. •It is possible to reduce the T-concentration below one DAC in TBM port cells. •The TBM port cells can have human access within 12 h after shutdown. -- Abstract: Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown.

  5. Singular point analysis during rail deployment into vacuum vessel for ITER blanket maintenance

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Shibanuma, Kiyoshi

    2007-05-01

    Remote maintenance of the ITER blanket composed of about 400 modules in the vessel is required by a maintenance robot due to high gamma radiation of ∼500Gy/h in the vessel. A concept of rail-mounted vehicle manipulator system has been developed to apply to the maintenance of the ITER blanket. The most critical issue of the vehicle manipulator system is the feasibility of the deployment of the articulated rail composed of eight rail links into the donut-shaped vessel without any driving mechanism in the rail. To solve this issue, a new driving mechanism and procedure for the rail deployment has been proposed, taking account of a repeated operation of the multi-rail links deployed in the same kinematical manner. The new driving mechanism, which is deferent from those of a usual 'articulated arm' equipped with actuator in the every joint for movement, is composed of three mechanisms. To assess the feasibility of the kinematics of the articulated rail for rail deployment, a kinematical model composed of three rail links related to a cycle of the repeated operation for rail deployment was considered. The determinant det J' of the Jacobian matrix J' was solved so as to estimate the existence of a singular point of the transformation during rail deployment. As a result, it is found that there is a singular point due to det J'=0. To avoid the singular point of the rail links, a new location of the second driving mechanism and the related rail deployment procedure are proposed. As a result of the rail deployment test based on the new proposal using a full-scale vehicle manipulator system, the respective rail links have been successfully deployed within 6 h less than the target of 8 h in the same manner of the repeated operation under a synchronized cooperation among the three driving mechanisms. It is therefore concluded that the feasibility of the rail deployment of the articulated rail composed of simple structures without any driving mechanism has been demonstrated

  6. 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.)

  7. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  8. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou

    1998-01-01

    ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

  9. Development of blanket remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

  10. Rail deployment and storage procedure and test for ITER blanket remote maintenance

    International Nuclear Information System (INIS)

    Kakudate, S.; Shibanuma, K.

    2003-01-01

    A concept of rail-mounted vehicle manipulator system has been developed to apply to the maintenance of the ITER blanket composed of ∼400 modules in the vacuum vessel. The most critical issue of the vehicle manipulator system is the feasibility of the deployment and storage of the articulated rail, composed of eight rail links without any driving mechanism in the joints. To solve this issue, a new driving mechanism and procedure for the rail deployment and storage has been proposed, taking account of the repeated operation of the multi-rail links deployed and stored in the same kinematical manner. The new driving mechanism, which is different from those of a usual articulated manipulator or 'articulated boom' equipped with actuators in every joint for movement, is composed of three external mechanisms installed outside the articulated rail, i.e. a vehicle traveling mechanism as main driver and two auxiliary driving mechanisms. A simplified synchronized control of three driving mechanisms has also been proposed, including 'torque-limit control' for suppression of the overload of the mechanisms. These proposals have been tested using a full-scale vehicle manipulator system, in order to demonstrate the proof of principle for rail deployment and storage. As a result, the articulated rail has been successfully deployed and stored within 6 h each, less than the target of 8 h, by means of the three external driving mechanisms and the proposed synchronized control. In addition, the overload caused by an unexpected mismatch of the synchronized control of three driving mechanisms has also been successfully suppressed less than the rated torque by the proposed 'torque-limit control'. It is therefore concluded that the feasibility of the rail deployment and storage of the vehicle manipulator system has been demonstrated

  11. Kinematic analysis on rail development into vacuum vessel for ITER blanket maintenance

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Shibanuma, Kiyoshi

    2006-01-01

    The vehicle manipulator system for blanket maintenance is used as a main driving mechanism for rail development, and three driving mechanisms d1, d2 (or d2') and d3 are used as cycle sequence of the repeated operations for rail development. This repeated operation can develop the articulated rail into the vacuum vessel. The rail development scenario, kinematic analysis model for rail development without any driving mechanisms in the rail joints, equations defined the angular between two rail links, identification of rail link at repeated operation, numerical analysis results on rail deployment under the forced position control of l i+1 , new rail development scenario using two driving mechanisms d1 and d2''under one cycle sequence of the repeated operations, and rail development test are reported. (S.Y.)

  12. 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)

  13. 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)

  14. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

    The present report summarizes the design activities of the ITER first wall and shielding blanket conducted by the JA Home Team during this year (1994) in close contact with the JCT, and reported during the four Technical Meetings held at Garching ITER Co-center. These activities are based on the Task Agreement between the JCT and the JA Home Team. In the present report, a layered configuration composed of separate first walls, modular-type blanket modules and separate back plates has been proposed to realize reliable assembly and maintenance schemes as well as to realize reliable component designs under high surface heat loads, high neutron wall loading and electromagnetic loads during disruptions. Outline of the structural design, consideration on fabricability and maintainability, and the results of thermal, mechanical and electromagnetic analyses are described. (author)

  15. 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

  16. Performance test of diamond-like carbon films for lubricating ITER blanket maintenance equipment under GPa-level high contact stress

    International Nuclear Information System (INIS)

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

    2007-01-01

    Diamond-like carbon (DLC) coating was tested as a candidate solid lubricant for transmission gears of the maintenance equipment of the blanket of the ITER instead of an oil lubricant. The wear tests using the pin-on-disk method were performed on disks with SCM440 and SNCM420 as the base materials and coated with soft, layered, and hard DLCs. All cases satisfied the required allowable contact stress (2 GPa) and lifetime (10 4 cycles), and therefore the feasibility of the DLC coating was validated. Among the three types of DLCs, the soft DLC showed the best performance. (author)

  17. 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)

  18. Construction of a test platform for Test Blanket Module (TBM) systems integration and maintenance in ITER Port Cell #16

    Energy Technology Data Exchange (ETDEWEB)

    Vála, Ladislav, E-mail: ladislav.vala@cvrez.cz [Centrum výzkumu Řež, Hlavní 130, 250 68 Husinec-Řež (Czech Republic); Reungoat, Mathieu, E-mail: mathieu.reungoat@cvrez.cz [Centrum výzkumu Řež, Hlavní 130, 250 68 Husinec-Řež (Czech Republic); Vician, Martin [Centrum výzkumu Řež, Hlavní 130, 250 68 Husinec-Řež (Czech Republic); Poitevin, Yves; Ricapito, Italo; Zmitko, Milan; Panayotov, Dobromir [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • A non-nuclear, full size facility – TBM platform – is under construction in CVR. • It is designed for tests, optimization and validation of TBS maintenance operations. • It will allow testing and validation of specific maintenance tools and RH equipment. • It reproduces ITER Port Cell #16, as well as the TBS interfaces and main equipment. • The TBM platform will be available for full operation in the first half of 2016. - Abstract: This paper describes a project of a non-nuclear, 1:1 scale testing platform dedicated to tests, optimization and validation of integration and maintenance operations for the European TBM systems in the ITER Port Cell #16. This TBM platform is currently under construction in Centrum výzkumu Řež, Czech Republic. The facility is realized within the scope of the SUSEN project and its full operation is foreseen in the first half of 2016.

  19. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

    1997-05-01

    A mechanical configuration of ITER integrated primary first wall/shield blanket module were developed focusing on the welded attachment of its support leg to the back plate. A 100 mm x 150 mm space between the legs of adjacent modules was incorporated for the working space of welding/cutting tools. A concept of coolant branch pipe connection to accommodate deformation due to the leg welding and differential displacement of the module and the manifold/back plate during operation was introduced. Two-dimensional FEM analyses showed that thermal stresses in Cu-alloy (first wall) and stainless steel (first wall coolant tube and shield block) satisfied the stress criteria following ASME code for ITER BPP operation. On the other hand, three-dimensional FEM analyses for overall in-vessel structures exhibited excessive primary stresses in the back plate and its support structure to the vacuum vessel under VDE disruption load and marginal stresses in the support leg of module No.4. Fabrication procedure of the integrated primary first wall/shield blanket module was developed based on single step solid HIP for the joining of Cu-alloy/Cu-alloy, Cu-alloy/stainless steel, and stainless steel/stainless steel. (author)

  20. Availability analysis of the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

    2015-01-01

    The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

  1. 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)

  2. Development of a control system for a heavy object handling manipulator. Application to a remote maintenance system for ITER blanket module

    International Nuclear Information System (INIS)

    Yoshimi, Takashi; Tsuji, Kouichi; Miyagawa, Shinichi; Kubo, Tomomi; Kakudate, Satoshi; Tada, Eisuke

    2001-01-01

    This paper describes a control system for the heavy object handling manipulator. It has been developed for the blanket module remote maintenance system of ITER (International Thermonuclear Fusion Experimental Reactor). A rail-mounted vehicle-type manipulator is proposed for the precise handling of a blanket module which is about 4 tons in weight. Basically, this manipulator is controlled by teaching-playback technique. When grasping or releasing the module, the manipulator sags and the position of the end-effector changes about 50 [mm]. Applying only the usual teaching-playback control makes the smooth operation of setting/removing modules to/from the vacuum vessel wall difficult due to this position change. To solve this proper problem of heavy object handling manipulator, we have developed a system which uses motion patterns generated from two kinds of teaching points. These motion patterns for setting/removing heavy objects are generated by combining teaching points for positioning the manipulator with and without grasping the object. When these motion patterns are applied, the manipulator can transfer the object's weight smoothly at the setting/removing point. This developed system has been applied to the real-scale mock-up of the vehicle manipulator and through the actual module setting/removing experiments, we have verified its effectiveness and realized smooth maintenance operation. (author)

  3. Design and analysis of ITER shield blanket

    Energy Technology Data Exchange (ETDEWEB)

    Ohmori, Junji; Hatano, Toshihisa; Ezato, Kouichiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-12-01

    This report includes electromagnetic analyses for ITER shielding blanket modules, fabrication methods for the blanket modules and the back plate, the design and the fabrication methods for port limiter have been investigated. Studies on the runaway electron impact for Be armor have been also performed. (J.P.N.)

  4. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  5. Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J., E-mail: Brad.Merrill@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Wong, C.P.C. [General Atomics, San Diego, CA 92186-5608 (United States); Cadwallader, L.C. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Abdou, M.; Morley, N.B. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)

    2014-10-15

    A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the {sup 210}Po and {sup 203}Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.

  6. 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.)

  7. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kuroda, Toshimasa; Enoeda, Mikio; Kikuchi, Shigeto [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-11-01

    The ITER breeding blanket employs a ceramic breeder and Be neutron multiplier both in small spherical pebble form. Radial-poloidal cooling panels are arranged in the blanket box to remove the nuclear heating in these materials and to reinforce the blanket structure. At the first wall, Be armor is bonded onto the stainless steel (SS) structure to provide a low Z plasma-compatible surface and to protect the first wall/blanket structure from the direct contact with the plasma during off-normal events. Thermo-mechanical analyses and investigation of fabrication procedure have been performed for this breeding blanket. To evaluate thermo-mechanical behavior of the pebble beds including the dependency of the effective thermal conductivity on stress, analysis methods have been preliminary established by the use of special calculation option of ABAQUS code, which are briefly summarized in this report. The structural response of the breeding blanket module under internal pressure of 4 MPa (in case of in-blanket LOCA) resulted in rather high stress in the blanket side (toroidal end) wall, thus addition of a stiffening rib or increase of the wall thickness will be needed. Two-dimensional elasto-plastic analyses have been performed for the Be/SS bonded interface at the first wall taking a fabrication process based on HIP bonding and thermal cycle due to pulsed plasma operation into account. The stress-strain hysteresis during these process and operation was clarified, and a procedure to assess and/or confirm the bonding integrity was also proposed. Fabrication sequence of the breeding blanket module was preliminarily developed based on the procedure to fabricate part by part and to assemble them one by one. (author)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  9. 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.

  10. Blanket maintenance by remote means using the cassette blanket approach

    International Nuclear Information System (INIS)

    Werner, R.W.

    1978-01-01

    Induced radioactivity in the blanket and other parts of a fusion reactor close to the plasma zone will dictate remote assembly, disassembly, and maintenance procedures. Time will be of the essence in these procedures. They must be practicable and certain. This paper discusses the reduction of a complicated Tokamak reactor to a simpler assembly via the use of a vacuum building in which to house the reactor and the introduction in this new model of cassette blanket modules. The cassettes significantly simplify remote handling

  11. NET test blanket design and remote maintenance

    International Nuclear Information System (INIS)

    Holloway, C.; Hubert, P.

    1991-01-01

    The NET machine has three horizontal ports reserved for testing tritium breeding blanket designs during the physics phase and possibly five during the technology phase. The design of the ports and test blankets are modular to accept a range of blanket options, provide radiation shielding and allow routine replacement. Radiation levels during replacement or maintenance require that all operations must be carried out remotely. The paper describes the problems overcome in providing a port design which includes attachment to the vacuum vessel with double vacuum seals, an integrated cooled first wall and support guides for the test blanket module. The method selected to remotely replace the test module whilst controlling the spread of contamination is also adressed. The paper concludes that the provisions of a test blanket facility based on the NET machine design is feasible. (orig.)

  12. Development of a virtual reality simulator for the ITER blanket remote handling system

    International Nuclear Information System (INIS)

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

    2008-01-01

    The authors developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robotic simulation software, ENVISION. The simulator is connected to the control system of the manipulator, which was developed as part of the blanket maintenance system during the Engineering Design Activity (EDA), and can reconstruct the positions of the manipulator and blanket module using position data transmitted from motors through a LAN. In addition, it can provide virtual visual information (e.g., about the interface structures behind the blanket module) by making the module transparent on the screen. It can also be used for confirming a maintenance sequence before the actual operation. The simulator will be modified further, with addition of other necessary functions, and will finally serve as a prototype of the actual simulator for the blanket remote handling system, which will be procured as part of an in-kind contribution

  13. Thermomechanical analysis of the DFLL test blanket module for ITER

    International Nuclear Information System (INIS)

    Chen Hongli; Wu Yican; Bai Yunqing

    2006-01-01

    The finite element code is used to simulate two kinds of blanket design structure, which are SLL (Quasi-Static Lithium Lead) and DLL (Dual-cooled Lithium Lead) blanket concepts for the Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM) submitted to the ITER test blanket working group. The temperature and stress distributions have been presented for the two kinds of blanket structure on the basis of the structural design, thermal-hydraulic design and neutronics analysis. Also the mechanical performance is presented for the high temperature component of blanket structure according to the ITER Structural Design Criteria (ISDC). The rationality and feasibility of the two kinds of blanket structure design of DFLL-TBM have been analyzed based on the above results which also acted as the theoretical base for further optimized analysis. (authors)

  14. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.

    1989-01-01

    The Li-particulate blanket design concept the authors proposed for the International Thermonuclear Experimental Reactor (ITER) uses a dilute suspension of fine solid breeder particles in a carrier gas as the combined coolant and lithium breeder stream. This blanket concept has a simple mechanical and hydraulic configuration, low inventory of bred tritium, and simple tritium extraction system. Existing technology can be used to implement the design for ITER. The concept has the potential to be a highly reliable shield and blanket design for ITER with relatively low development and capital costs

  15. Electrical connectors for blanket modules in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I., E-mail: poddubnyyii@nikiet.ru [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Khomiakov, S.; Kolganov, V. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Sadakov, S.; Calcagno, B.; Chappuis, Ph.; Roccella, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Danilov, I.; Leshukov, A.; Strebkov, Y. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Ulrickson, M. [Sandia National Laboratories MS-1129, PO Box 5800, Albuquerque, NM 87185 (United States)

    2014-10-15

    Highlights: • Analysis of static and cyclic strength for L-shaped and Z-shaped ES has been performed. • Analysis results do show that for L-shaped ES static and cyclic strength criteria are not satisfied. • Static and cyclic strength criteria are met well by ES with Z-shaped elastic elements. • ES with Z-shaped elastic elements has been adopted as a new baseline design for ITER. - Abstract: Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak. The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was

  16. Measurement and control system for ITER remote maintenance equipment

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Kakudate, Satoshi; Takeda, Nobukazu; Takiguchi, Yuji; Akou, Kentaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    ITER in-vessel components such as blankets and divertors are categorized as scheduled maintenance components because they are subjected to severe plasma heat and particle loads. Blanket maintenance requires remote handling equipment and tools able to handle Heavy payloads of about 4 tons within a 2 mm precision tolerance. Divertor maintenance requires remote replacement of 60 cassettes with a dead weight of about 25 tons each. In the ITER R and D program, full-scale remote handling equipment for blanket and divertor maintenance has been designed and assembled for demonstration tests. This paper reviews the measurement and control system developed for full-scale remote handling equipment, the Japan Home Team contribution. (author)

  17. Measurement and control system for ITER remote maintenance equipment

    International Nuclear Information System (INIS)

    Oka, Kiyoshi; Kakudate, Satoshi; Takeda, Nobukazu; Takiguchi, Yuji; Akou, Kentaro

    1998-01-01

    ITER in-vessel components such as blankets and divertors are categorized as scheduled maintenance components because they are subjected to severe plasma heat and particle loads. Blanket maintenance requires remote handling equipment and tools able to handle Heavy payloads of about 4 tons within a 2 mm precision tolerance. Divertor maintenance requires remote replacement of 60 cassettes with a dead weight of about 25 tons each. In the ITER R and D program, full-scale remote handling equipment for blanket and divertor maintenance has been designed and assembled for demonstration tests. This paper reviews the measurement and control system developed for full-scale remote handling equipment, the Japan Home Team contribution. (author)

  18. Electromagnetic analysis of ITER shield blanket under VDE

    International Nuclear Information System (INIS)

    Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

    2010-01-01

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

  19. 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

  20. 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.

  1. Status of blanket design for RTO/RC ITER

    International Nuclear Information System (INIS)

    Yamada, M.; Ioki, K.; Cardella, A.; Elio, F.; Miki, N.

    2000-01-01

    Design has progressed on the FW/blanket for the RTO/RC (reduced technical objective/ reduced cost) ITER. The basic functions and structures are the same as for the 1998 ITER design. However, design and fabrication methods of the FW/blanket have been improved to achieve ∝ 50% reduction of the construction cost compared to that for the 1998 ITER design. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the EDA (engineering design activity) is still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed. (orig.)

  2. An assessment of the base blanket for ITER

    International Nuclear Information System (INIS)

    Raffray, A.R.; Abdou, M.A.; Ying, A.

    1991-01-01

    Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored

  3. 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.)

  4. 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)

  5. ITER solid breeder blanket materials database

    International Nuclear Information System (INIS)

    Billone, M.C.; Dienst, W.; Noda, K.; Roux, N.

    1993-11-01

    The databases for solid breeder ceramics (Li 2 ,O, Li 4 SiO 4 , Li 2 ZrO 3 and LiAlO 2 ) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized

  6. ITER blanket module shield block design and analysis

    International Nuclear Information System (INIS)

    Mitin, D.; Khomyakov, S.; Razmerov, A.; Strebkov, Yu.

    2008-01-01

    This paper presents the alternative design of the shield block cooling path for a typical ITER blanket module with a predominantly sequential flow circuit. A number of serious disadvantages have been observed for the reference design, where the parallel flow circuit is used, which is inherent in the majority of blanket modules. The paper discusses these disadvantages and demonstrates the benefit of the alternative design based on the detailed design and the technological, hydraulic, thermal, structural and strength analyses, conducted for module no. 17

  7. 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)

  8. 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)

  9. Fabrication of prototype mockups of ITER shielding blanket with separable first wall

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Akiba, Masato

    2002-07-01

    Design of shielding blanket for ITER-FEAT applies the first wall which has the separable structure from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. This report summarizes the demonstrative fabrication of the ITER shielding blanket with separable first wall performed for the shielding blanket fabrication technology development, under the task agreement of G 16 TT 108 FJ (T420-2) in ITER Engineering Design Activity Extension Period. The objectives of the demonstrative fabrication are: to demonstrate the comprehensive fabrication technique in a large scale component (e.g the joining techniques for the beryllium armor/copper alloy and copper alloy/SS, and the slotting method of the FW and shield block); to develop an improved fabrication method for the shielding blanket based on the ITER-FEAT updated design. In this work, the fabrication technique of full scale separable first wall shield blanket was confirmed by fabricating full width Be armored first wall panel, full scale of 1/2 shield block with poloidal cooling channels. As the R and D for updated cooling channel configuration, the fabrication technique of the radial channel shield block was also demonstrated. Concluding to the all R and D results, it was demonstrated successfully that the fabrication technique and optimized conditions in the results obtained under the task agreement of G 16 TT 95 FJ (T420-1) was applicable to the prototype of the separable first wall blanket module. Additionally, basic echo data of ultra-sonic test method (UT) was obtained to show the applicability of UT method for in tube access detection of defect on the Cu alloy/SS tube interface. (author)

  10. Radwaste management aspects of the test blanket systems in ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  11. Nuclear maintenance strategy and first steps for preliminary maintenance plan of the EU HCLL & HCPB Test Blanket Systems

    Energy Technology Data Exchange (ETDEWEB)

    Galabert, Jose, E-mail: jose.galabert@f4e.europa.eu [F4E Fusion for Energy, EU Domestic Agency, c/Josep Pla, 2. B3, 08019, Barcelona (Spain); Hopper, Dave [AMEC Foster Wheeler, Faraday Street, Birchwood Park, WA3 6GN (United Kingdom); Neviere, Jean-Cristophe [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, 13067, St. Paul Lez Durance Cedex (France); Nodwell, David [CCFE, Culham Science Centre, Abingdon, OX14 3DB, Oxfordshire (United Kingdom); Pascal, Romain [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, 13067, St. Paul Lez Durance Cedex (France); Poitevin, Yves; Ricapito, Italo [F4E Fusion for Energy, EU Domestic Agency, c/Josep Pla, 2. B3, 08019, Barcelona (Spain); White, Gareth [AMEC Foster Wheeler, Faraday Street, Birchwood Park, WA3 6GN (United Kingdom)

    2017-03-15

    Highlights: • Nuclear maintenance strategy for the two European (EU) Test Blanket Systems (TBS): i/. Helium Cooled Lead Lithium (HCLL) and ii/. Helium Cooled Pebble Bed (HCPB). • Preliminary identification of maintenance tasks for most relevant components of the EU HCLL & HCPB TBS. • Preliminary feasibility analysis for hands-on maintenance tasks of some relevant components of the European Test Blanket Systems. • Design recommendations for enhancement of the European Test Blanket Systems maintainability. - Abstract: This paper gives an overview of nuclear maintenance strategy to be followed for the European HCLL & HCPB Test Blanket Systems (TBS) to be installed in ITER. One of the several core documents to prepare in view of their licensing is their respective ‘Maintenance Plan’. This document is fundamental for ensuring sound performance and safety of the TBS during ITER’s operational phase and shall include, amongst others, relevant information on: maintenance organization, preventive and corrective maintenance task procedures, condition monitoring for key components, maintenance work planning, and a spare parts plan, just to mention some of the key topics. In compliance with the ITER Plant Maintenance policy, first steps have been taken aimed at defining nuclear maintenance strategy for some of the most relevant HCLL & HCPB TBS components, conducted by F4E in collaboration with industry. After a brief recall of maintenance strategy of the TBM Program (PBS-56), this paper analyses main features of EU HCLL & HCPB TBS maintainability and identifies, at their conceptual design phase, a preliminary list of maintenance tasks to be developed for their most representative components. In addition, the paper also presents the first nuclear maintenance studies conducted for replacement of the Q{sub 2} Getter Beds, identifying some design recommendations for their sound maintainability.

  12. Nuclear maintenance strategy and first steps for preliminary maintenance plan of the EU HCLL & HCPB Test Blanket Systems

    International Nuclear Information System (INIS)

    Galabert, Jose; Hopper, Dave; Neviere, Jean-Cristophe; Nodwell, David; Pascal, Romain; Poitevin, Yves; Ricapito, Italo; White, Gareth

    2017-01-01

    Highlights: • Nuclear maintenance strategy for the two European (EU) Test Blanket Systems (TBS): i/. Helium Cooled Lead Lithium (HCLL) and ii/. Helium Cooled Pebble Bed (HCPB). • Preliminary identification of maintenance tasks for most relevant components of the EU HCLL & HCPB TBS. • Preliminary feasibility analysis for hands-on maintenance tasks of some relevant components of the European Test Blanket Systems. • Design recommendations for enhancement of the European Test Blanket Systems maintainability. - Abstract: This paper gives an overview of nuclear maintenance strategy to be followed for the European HCLL & HCPB Test Blanket Systems (TBS) to be installed in ITER. One of the several core documents to prepare in view of their licensing is their respective ‘Maintenance Plan’. This document is fundamental for ensuring sound performance and safety of the TBS during ITER’s operational phase and shall include, amongst others, relevant information on: maintenance organization, preventive and corrective maintenance task procedures, condition monitoring for key components, maintenance work planning, and a spare parts plan, just to mention some of the key topics. In compliance with the ITER Plant Maintenance policy, first steps have been taken aimed at defining nuclear maintenance strategy for some of the most relevant HCLL & HCPB TBS components, conducted by F4E in collaboration with industry. After a brief recall of maintenance strategy of the TBM Program (PBS-56), this paper analyses main features of EU HCLL & HCPB TBS maintainability and identifies, at their conceptual design phase, a preliminary list of maintenance tasks to be developed for their most representative components. In addition, the paper also presents the first nuclear maintenance studies conducted for replacement of the Q_2 Getter Beds, identifying some design recommendations for their sound maintainability.

  13. Assessment of alkali metal coolants for the ITER blanket

    International Nuclear Information System (INIS)

    Natesan, K.; Reed, C.B.; Mattas, R.F.

    1994-01-01

    The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper will address the thermodynamics of interactions between the liquid metals (i.e., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data will be used to assess the long-term performance of the first wall in a liquid metal environment

  14. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  16. 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.

  17. Performance evaluation on force control for ITER blanket installation

    Energy Technology Data Exchange (ETDEWEB)

    Aburadani, A., E-mail: aburadani.atsushi@jaea.go.jp [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Nakahira, M.; Hamilton, D.; Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation.

  18. Performance evaluation on force control for ITER blanket installation

    International Nuclear Information System (INIS)

    Aburadani, A.; Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S.; Nakahira, M.; Hamilton, D.; Tesini, A.

    2013-01-01

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  1. Sensitivity and uncertainty analysis of NET/ITER shielding blankets

    International Nuclear Information System (INIS)

    Hogenbirk, A.; Gruppelaar, H.; Verschuur, K.A.

    1990-09-01

    Results are presented of sensitivity and uncertainty calculations based upon the European fusion file (EFF-1). The effect of uncertainties in Fe, Cr and Ni cross sections on the nuclear heating in the coils of a NET/ITER shielding blanket has been studied. The analysis has been performed for the total cross section as well as partial cross sections. The correct expression for the sensitivity profile was used, including the gain term. The resulting uncertainty in the nuclear heating lies between 10 and 20 per cent. (author). 18 refs.; 2 figs.; 2 tabs

  2. The ITER EC H and CD Upper Launcher: Maintenance concepts

    International Nuclear Information System (INIS)

    Ronden, D.M.S.; Baar, M. de; Chavan, R.; Elzendoorn, B.S.Q.; Grossetti, G.; Heemskerk, C.J.M.; Koning, J.F.; Landis, J.-D.; Spaeh, P.; Strauss, D.

    2013-01-01

    Highlights: ► We explain how an overall maintenance strategy defines individual maintenance tasks. ► Concepts are presented for replacement strategies of the in-vessel optical components. ► Vertical placement of the Upper Launcher in the Hot Cell may simplify maintenance. -- Abstract: Maintenance of the ITER EC H and CD Upper Launcher (UL) shall be performed through the use of Remote Handling (RH) in the ITER Hot Cell Facility (HCF). The UL design will have to be fully compliant with ITER RH maintenance requirements and the set of RH tooling and services available in the HCF. This paper describes the development of an overall maintenance strategy for the UL, starting from a listing of all conceivable maintenance operations, including hands-on tasks. Components for which design concepts are discussed in this paper are the Blanket Shield Module (BSM), the steering mirror (M4), the mid optics (M1, M2) and the waveguide (WG) feed-through plate. Aspects related to RH documentation, overall maintenance strategy and design concepts for optimizing the maintainability of the UL are presented

  3. Tritium inventory and permeation in the ITER breeding blanket

    International Nuclear Information System (INIS)

    Violante, V.; Tosti, S.; Sibilia, C.; Felli, F.; Casadio, S.; Alvani, C.

    2000-01-01

    A model has allowed us to perform the analysis of the tritium inventory and permeation in the international thermonuclear experimental reactor (ITER) breeding blanket under the hypothesis of steady state conditions. Li 2 ZrO 3 (reference) and Li 2 TiO 3 (alternative) have been studied as breeding materials. The total breeder inventory assessed is 7.64 g for the Li 2 ZrO 3 at reference temperature. The model has also been used for a parametric analysis of the tritium permeation. At reference temperature and purge helium velocity of 0.01 m/s, the HT partial pressure is ranging from 10 to 30 Pa in the breeder and 1.5x10 -3 Pa in the beryllium. At 0.1 m/s of purge helium velocity, the HT partial pressure is reduced of one order by magnitude in the breeder and becomes 5x10 -5 Pa in the beryllium. The tritium permeation into the coolant for the whole blanket is ranging from 100 to 250 mCi per day for purge helium velocity of 0.01 m/s. The analysis of the tritium inventory and permeation for the alternative Li 2 TiO 3 breeding material has been carried out too. The tritium inventory in the breeder is in the range from 6 to 375 g larger than in Li 2 ZrO 3 by about a factor 5; the tritium permeation into coolant is comparable to the Li 2 ZrO 3 one. This analysis provides indications on the influence of the operating parameters on the tritium control in the ITER breeding blanket; particularly the control of the tritium inventory by the temperature and the tritium permeation by the purge gas velocity

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

    International Nuclear Information System (INIS)

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

    2006-08-01

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

  5. Fabrication of the full scale separable first wall of ITER shielding blanket

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Hatano, Toshihisa; Enoeda, Mikio; Miki, Nobuharu; Akiba, Masato

    2002-10-01

    Shielding blanket for ITER-FEAT applies the unique first wall structure which is separable from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. Such unique features of blanket structure required technological clarification from the technical base of the previous achievement of the blanket module fabrication development. Previously, within the EDA Task T216+, a prototype for the no.4 Primary Wall Module of the ITER Shield Blanket with integrated first wall has been manufactured by forging and drilling and the first wall has been manufactured and joined to the shield block by Hot Isostatic Pressing (HIP) in one step process. This work has been performed to clarify the remaining R and D issues which have not been covered in the previous R and D. This report summarizes the demonstrative fabrication of the real scale separable first wall for ITER shielding blanket designed for ITER-FEAT, together with the essential technology developments such as, the slit grooving of the first wall with beryllium armor and SS shield block and fabrication of a partial mockup of beryllium armored first wall panel with built-in cooling channels. This work has been performed under the task agreement of G 16 TT 95 FJ (T420-1) in ITER Engineering Design Activity Extension Period. By the demonstration of the Be armor joining to the first wall panel, the joining technique of Be and DSCu developed previously, was shown to be applicable to the realistic structure of first wall panel. Also, the slit grooving by an end-mill method and an electron discharge machining method have been applied to the first wall mockup with Be armor tiles and demonstrated the applicability within the design tolerance. As the slit grooving technique

  6. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-11-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

  7. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

    2001-01-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

  8. Materials development for ITER shielding and test blanket in China

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

    2011-10-01

    China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

  9. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

  10. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    International Nuclear Information System (INIS)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li 2 O) and lithium zirconate (Li 2 ZrO 3 ) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers

  11. 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.

  12. Development of ITER shielding blanket prototype mockup by HIP bonding

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi; Furuya, Kazuyuki; Hatano, Toshihisa; Kuroda, Toshimasa; Enoeda, Mikio; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Takatsu, Hideyuki [Japan Atomic Energy Research Inst., Office of ITER Project Promotion, Tokyo (Japan)

    2000-07-01

    A prototype ({approx}900{sup H} x 1700{sup W} x 350{sup T} mm) of the ITER shielding blanket module has been fabricated following the previous successful fabrication of a small-scale ({approx}500{sup H} x 400{sup W} x 150{sup T} mm) and mid-scale ({approx}800{sup H} x 500{sup W} x 350{sup T} mm) mock-ups. This prototype incorporates most of key design features essential to the fabrication of the ITER shielding blanket module such as 1) the first wall heat sink made of Al{sub 2}O{sub 3} dispersion strengthened Cu (DSCu) with built-in SS316L coolant tubes bonded to a massive SS316LN shield block, 2) toroidally curved first wall with a radius of 5106 mm while straight in poloidal direction, 3) coolant channels oriented in poloidal direction in the first wall and in toroidal direction in the shield block, 4) the first wall coolant channel routing to avoid the interference with the front access holes, 5) coolant channels drilled through the forged SS316LN-IG shield block, and 6) four front access holes of 30 mm in diameter penetrated through the first wall and the shield block. For the joining method, especially for the first wall/side wall parts and the shield block, the solid HIP (Hot Isostatic Pressing) process was applied. It is difficult to apply conventional joining methods such as field welding, brazing, explosion bonding and mechanical one-axial diffusion bonding to a wide area bonding because sufficient mechanical strengths can not be obtained and excessive deformations occurs. In order to solve these fabrication issues, HIP bonding was applied. The first wall stainless steel (SS) coolant tubes of 10 mm in inner diameter and l mm in thickness were sandwiched by semi-circular grooved DSCu plates at the first wall and the front region of the side wall, and by semi-circular grooved SS plates at the back region of the side wall. After assembling of these first wall/side wall parts with the shield block, they were simultaneously bonded by single step HIP in order to

  13. Assessment of alternative vessel and blanket design on ITER operation

    Energy Technology Data Exchange (ETDEWEB)

    Cavinato, M., E-mail: mario.cavinato@f4e.europa.e [FUSION FOR ENERGY Joint Undertaking, 08019 Barcelona (Spain); Portone, A.; Saibene, G.; Sartori, R. [FUSION FOR ENERGY Joint Undertaking, 08019 Barcelona (Spain); Albanese, R.; Ambrosino, G.; Ariola, M. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Napoli (Italy); Artaserse, G. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Reggio Calabria (Italy); Mattei, M. [Associazione Euratom-ENEA-CREATE, DIAM, Seconda Universita di Napoli, Via Roma 29, Aversa, CE 81031 Italy (Italy); Pironti, A. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Napoli (Italy); Villone, F. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Cassino (Italy)

    2010-12-15

    In the framework of the ITER project, an investigation has been conducted on an alternative vessel and blanket design, aimed at reducing cost and production risk. The modifications proposed have a strong impact on plasma control since they affect the main conducting structures surrounding the plasma column, providing passive stabilization but at the same time shielding the field generated by the active coils to control the plasma motion and shape. An extensive analysis was performed to assess the plasma vertical controllability and the modified requirements to the in-vessel vertical stability coils system as well as to the external Poloidal Field coils system. A similar analysis was aimed at assessing the performance of the shape control system in presence of the modified structures. The effect on plasma breakdown was also evaluated in terms of maximum initial loop voltage, quality of magnetic null and the flux loss related to the breakdown delay that was quantified under the same hypothesis employed by ITER for the baseline design. Furthermore, the modified design presents issues for the magnetic diagnostic system, related to the shielding of the probes by the eddy currents, which were analysed with a 3D model. The results of the analyses performed have some general interest in particular regarding the influence on plasma stability of 3D structures with close proximity to the plasma. The present paper aims at giving an overview of the analyses that have been carried out and a summary of the results in terms of impact of the modified design on plasma control and scenario, and in general an evaluation of the role of passive structure in plasma vertical stability and shape control.

  14. Analysis of ITER upper port plug remote handling maintenance scenarios

    International Nuclear Information System (INIS)

    Koning, J.F.; Baar, M.R. de; Elzendoorn, B.S.Q.; Heemskerk, C.J.M.; Ronden, D.M.S.; Schuth, W.J.

    2012-01-01

    Highlights: ► Remote Handling Study Centre: providing RH compatibility analysis. ► Simulation: virtual reality including kinematics and realtime physics simulator. ► Applied on analysis of RH compatibility of Upper Launcher component replacement. ► Resulting in lowered maintenance procedure time and lessons learned. - Abstract: The ITER tokamak has a modular design, with port plugs, blanket modules and divertor cassettes. This set-up allows for maintenance of diagnostics, heating systems and first wall elements. The maintenance can be done in situ, or in the Hot Cell. Safe and effective remote handling (RH) will be ensured by the RH requirements and standards. Compliance is verified through remote handling compatibility assessments at the ITER Design Review milestones. The Remote Handling Study Centre at FOM Institute DIFFER is created to study ITER RH maintenance processes at different levels of complexity, from relatively simple situational awareness checks using snap-shots in the CAD system, time studies using virtual reality (VR) animations, to extensive operational sequence validation with multiple operators in real-time. The multi-operator facility mimics an RH work-cell as presently foreseen in the ITER RH control room. Novel VR technology is used to create a realistic setting in which a team of RH operators can interact with virtual ITER environments. A physics engine is used to emulate real-time contact interaction as to provide realistic haptic feed-back. Complex interactions between the RH operators and the control room system software are tested. RH task performance is quantified and operational resource usage estimated. The article provides a description and lessons learned from a recent study on replacement of the Steering Mirror Assembly on the ECRH (Electron Cyclotron Resonance Heating) Upper Launcher port plug.

  15. Analysis of ITER upper port plug remote handling maintenance scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Koning, J.F., E-mail: j.f.koning@heemskerk-innovative.nl [FOM Institute DIFFER - Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Baar, M.R. de; Elzendoorn, B.S.Q. [FOM Institute DIFFER - Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Heemskerk, C.J.M. [Heemskerk Innovative Technology, Noordwijk (Netherlands); Ronden, D.M.S.; Schuth, W.J. [FOM Institute DIFFER - Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Remote Handling Study Centre: providing RH compatibility analysis. Black-Right-Pointing-Pointer Simulation: virtual reality including kinematics and realtime physics simulator. Black-Right-Pointing-Pointer Applied on analysis of RH compatibility of Upper Launcher component replacement. Black-Right-Pointing-Pointer Resulting in lowered maintenance procedure time and lessons learned. - Abstract: The ITER tokamak has a modular design, with port plugs, blanket modules and divertor cassettes. This set-up allows for maintenance of diagnostics, heating systems and first wall elements. The maintenance can be done in situ, or in the Hot Cell. Safe and effective remote handling (RH) will be ensured by the RH requirements and standards. Compliance is verified through remote handling compatibility assessments at the ITER Design Review milestones. The Remote Handling Study Centre at FOM Institute DIFFER is created to study ITER RH maintenance processes at different levels of complexity, from relatively simple situational awareness checks using snap-shots in the CAD system, time studies using virtual reality (VR) animations, to extensive operational sequence validation with multiple operators in real-time. The multi-operator facility mimics an RH work-cell as presently foreseen in the ITER RH control room. Novel VR technology is used to create a realistic setting in which a team of RH operators can interact with virtual ITER environments. A physics engine is used to emulate real-time contact interaction as to provide realistic haptic feed-back. Complex interactions between the RH operators and the control room system software are tested. RH task performance is quantified and operational resource usage estimated. The article provides a description and lessons learned from a recent study on replacement of the Steering Mirror Assembly on the ECRH (Electron Cyclotron Resonance Heating) Upper Launcher port plug.

  16. The ITER Remote Maintenance Management System

    International Nuclear Information System (INIS)

    Tesini, Alessandro; Rolfe, A.C.

    2009-01-01

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

  17. Materials issues in the design of the ITER first wall, blanket, and divertor

    International Nuclear Information System (INIS)

    Mattas, R.F.; Smith, D.L.; Wu, C.H.; Shatalov, G.

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R ampersand D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented

  18. Development of Joining Technologies for the ITER Blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byoung Kwon; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan; Lee, Dong Won; Kim, Suk Kwon [KAERI, Daejeon (Korea, Republic of)

    2011-01-15

    The design of the ITER blanket first wall includes the Beryllium amour tiles joined to CuCrZr heat sink with stainless steel cooling tubes. For the ITER application, the Be/CuCrZr/SS joint was proposed as a first wall material. The joining of Be/CuCrZr as well as CuCrZr/SS was generally carried out by using a hot isostatic pressing (CuC) in many countries. The joining strength for Be/CuCrZr is relatively lower than that for CuCrZr/SS, since we usually forms surface oxides (BeO) and brittle a metallics with Cu. Therefore, the joining technology for the Be/CuCrZr joint has been investigated. Be is apt to adsorb oxygen in an air atmosphere, so we should be etched to eliminate the surface pre-oxide using a chemical solution and Ar ions in a vacuum chamber. Then we is coated with a first was to prevent further oxidation. The kinds of a first we are chosen to be able to enhance the joining strength as inhibiting excessive be diffusion. The performance of the Be/CuCrZr/SS joint used for the ITER first wall is primarily dependent on the joining strength of the Be/CuCrZr interface. The Cr/Cu and Ti/Cr/Cu interlayers enabled the successful joining of be tile to CuCrZr plate. Moreover, ion-beam assisted deposition (IBAD) increased joining strength of the Be/CuCrZr joint mock-ups. IBAD induced the increased packing of depositing atoms, which resulted in denser and more adhesive interlayers. The interlayers formed by IBAD process revealed about 40% improved resistance to the scratch test. It is suggested that the improved adhesion of coating interlayers enabled tight joining of Be and CuCrZr blocks. As compared to without IBAD coating, the shear strength as well as the 4-point bend strength were increased more than 20% depending on interlayer types and coating conditions

  19. Development of Joining Technologies for the ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Choi, Byoung Kwon; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan; Lee, Dong Won; Kim, Suk Kwon

    2011-01-01

    The design of the ITER blanket first wall includes the Beryllium amour tiles joined to CuCrZr heat sink with stainless steel cooling tubes. For the ITER application, the Be/CuCrZr/SS joint was proposed as a first wall material. The joining of Be/CuCrZr as well as CuCrZr/SS was generally carried out by using a hot isostatic pressing (CuC) in many countries. The joining strength for Be/CuCrZr is relatively lower than that for CuCrZr/SS, since we usually forms surface oxides (BeO) and brittle a metallics with Cu. Therefore, the joining technology for the Be/CuCrZr joint has been investigated. Be is apt to adsorb oxygen in an air atmosphere, so we should be etched to eliminate the surface pre-oxide using a chemical solution and Ar ions in a vacuum chamber. Then we is coated with a first was to prevent further oxidation. The kinds of a first we are chosen to be able to enhance the joining strength as inhibiting excessive be diffusion. The performance of the Be/CuCrZr/SS joint used for the ITER first wall is primarily dependent on the joining strength of the Be/CuCrZr interface. The Cr/Cu and Ti/Cr/Cu interlayers enabled the successful joining of be tile to CuCrZr plate. Moreover, ion-beam assisted deposition (IBAD) increased joining strength of the Be/CuCrZr joint mock-ups. IBAD induced the increased packing of depositing atoms, which resulted in denser and more adhesive interlayers. The interlayers formed by IBAD process revealed about 40% improved resistance to the scratch test. It is suggested that the improved adhesion of coating interlayers enabled tight joining of Be and CuCrZr blocks. As compared to without IBAD coating, the shear strength as well as the 4-point bend strength were increased more than 20% depending on interlayer types and coating conditions

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-15

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

  1. Breeding blanket design for ITER and prototype (DEMO) fusion reactors and breeding materials issues

    Energy Technology Data Exchange (ETDEWEB)

    Takatsu, H; Enoeda, M [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    Current status of the designs of the ITER breeding blanket and DEMO blankets is introduced placing emphasis on the breeding materials selection and related issues. The former design is based on the up-to-date design activities, as of October 1997, being performed jointly by Joint Central Team (JCT) and Home Teams (HT`s), while the latter is based on the DEMO blanket test module designs being proposed by each Party at the TBWG (Test Blanket Working Group) meetings. (J.P.N.)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  3. Japanese contributions to ITER testing program of solid breeder blankets for DEMO

    International Nuclear Information System (INIS)

    Kuroda, Toshimasa; Yoshida, Hiroshi; Takatsu, Hideyuki; Maki, Koichi; Mori, Seiji; Kobayashi, Takeshi; Suzuki, Tatsushi; Hirata, Shingo; Miura, Hidenori.

    1991-04-01

    ITER Conceptual Design Activity (CDA), which has been conducted by four parties (Japan, EC, USA and USSR) since May 1988, has been finished on December 1990 with a great achievement of international design work of the integrated fusion experimental reactor. Numerous issues of physics and technology have been clarified for providing a framework of the next phase of ITER (Engineering Design Activity; EDA). Establishment of an ITER testing program, which includes technical test issues of neutronics, solid breeder blankets, liquid breeder blankets, plasma facing components, and materials, has been one of the goals of the CDA. This report describes Japanese proposal for the testing program of DEMO/power reactor blanket development. For two concepts of solid breeder blanket (helium-cooled and water-cooled), identification of technical issues, scheduling of test program, and conceptual design of test modules including required test facility such as cooling and tritium recovery systems have been carried out as the Japanese contribution to the CDA. (author)

  4. ITER Remote Maintenance System (IRMS) lifecycle management

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

  5. ITER Remote Maintenance System (IRMS) lifecycle management

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. Pulsed activation analyses of the ITER blanket design options considered in the blanket trade-off study

    International Nuclear Information System (INIS)

    Wang, Q.; Henderson, D.L.

    1995-01-01

    Pulsed activation calculations have been performed on two blanket options considered as part of the ITER home team blanket trade-off study. The objective was to compare the activity, afterheat and waste disposal rating (WDR) results of a composite blanket-shield design for the continuous operation approximation to a pulsed operation case to determine whether the differences are at most the duty factor as predicted by the two nuclide chain model. Up to a cooling period of 100 years, the pulsed activity and afterheat values were below the continuous oepration results and well within (except for one afterheat value) the maximum deviation predicted by the two nuclide chain model. No differences in the WDR values were noted as they are, to a large extent, based on long-lived nuclides which are insensitive to short-term changes in the operation history. (orig.)

  7. DEMO relevance of the test blanket modules in ITER-Application to the European test blanket modules

    International Nuclear Information System (INIS)

    Magnani, E.; Gabriel, F.; Boccaccini, L.V.; Li-Puma, A.

    2010-01-01

    Test blanket module (TBM) testing programme in ITER as a support to DEMO design is a very important step on the road map to commercial fusion reactors although it is an ambitious task. Finding as much as possible DEMO relevant tests in view of the future DEMO blanket design is therefore a major goal since ITER and DEMO environment and loading conditions are different. To clarify and quantify the meaning of the DEMO relevance, criteria using a structural, functional and behavioural representation of the breeding blanket acting as a system are investigated. Then, a three-step strategy is proposed to carry out TBM DEMO relevant tests associated with a TBM design modification strategy. Key parameters should intensively be used as target for TBM characterization and numerical code validation. When assessing the relevance, on the other hand, not only the actual difference between DEMO and ITER values should be considered, but also whether the analyzed phenomena have a threshold and a range of applicability, as numerical simulations are usually permitted within these limits. The proposed methodology is at the end applied to the design of the HCLL TBM breeding unit configuration.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

  10. Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2006-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  11. Monte Carlo analysis of helium production in the ITER shielding blanket module

    International Nuclear Information System (INIS)

    Sato, S.

    1999-01-01

    In order to examine the shielding performances of the inboard blanket module in the international thermonuclear experimental reactor (ITER), shielding calculations have been carried out using a three-dimensional Monte Carlo method. The impact of radiation streaming through the front access holes and gaps between adjacent blanket modules on the helium gas production in the branch pipe weld locations and back plate have been estimated. The three-dimensional model represents an 18 sector of the overall torus region and includes the vacuum vessel, inboard blanket and back plate, plasma region, and outboard reflecting medium. And it includes the 1 m high inboard mid-plane module and the 20 mm wide gaps between adjacent modules. From the calculated results for the reference design, it has been found that the helium production at the plug of the branch pipe is four to five times higher than the design goal of 1 appm for a neutron fluence of 0.9 MW a m -2 at the inboard mid-plane first wall. Also, it has been found that the helium production at the back plate behind the horizontal gap is about three times higher than the design goal. In the reference design, the stainless steel (SS):H 2 O composition in the blanket module is 80:20%. Shielding calculations also have been carried out for the SS:H 2 O composition of 70:30, 60:40, 50:50 and 40:60%. From the evaluated results for their design, it has been found that the dependence of helium production on the SS:H 2 170 mm will reduce helium production to satisfy the design goal and not have a significant impact on weight limitations imposed by remote maintenance handling limitations. Also based on the calculated results, about 200 mm thick shields such as a key structure in the vertical gap are suggested to be installed in the horizontal gap as well to reduce the helium production at the back plate and to satisfy the design goal. (orig.)

  12. 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)

  13. Liquid metal blanket module testing and design for ITER/TIBER II

    International Nuclear Information System (INIS)

    Mattas, R.F.; Cha, Y.; Finn, P.A.; Majumdar, S.; Picologlou, B.; Stevens, H.; Turner, L.

    1988-05-01

    A major goal for ITER is the testing of nuclear components to demonstrate the integrated performance of the most attractive concepts that can lead to a commercial fusion reactor. As part of the ITER/TIBER II study, the test program and design of test models were examined for a number of blanket concepts. The work at Argonne National Laboratory focused on self-cooled liquid metal blankets. A test program for liquid metal blankets was developed based upon the ITER/TIBER II operating schedule and the specific data needs to resolve the key issues for liquid metals. Testing can begin early in reactor operation with liquid metal MHD tests to confirm predictive capability. Combined heat transfer/MHD tests can be performed during initial plasma operation. After acceptable heat transfer performance is verified, tests to determine the integrated high temperature performance in a neutron environment can begin. During the high availability phase operation, long term performance and reliability tests will be performed. It is envisioned that a companion test program will be conducted outside ITER to determine behavior under severe accident conditions and upper performance limits. A detailed design of a liquid metal test module and auxiliary equipment was also developed. The module followed the design of the TPSS blanket. Detailed analysis of the heat transfer and tritium systems were performed, and the overall layout of the systems was determined. In general, the blanket module appears to be capable of addressing most of the testing needs. 8 refs., 27 figs., 11 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  15. 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

  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 [International Thermonuclear Experimental Reactor] shield and blanket work package report

    International Nuclear Information System (INIS)

    1988-06-01

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs

  18. Improved modules for the blanket of RTO/RC ITER

    International Nuclear Information System (INIS)

    Elio, F.; Ioki, K.; Cardella, A.

    2000-01-01

    This paper describes innovative design aspects that are considered to optimise the blanket modules for the reduced technical objective/reduced cost international thermonuclear experimental reactor. The blanket modules have a vertical straight profile facing the plasma, and the first wall is built in small and flat panels. Copper may be applied only in front of the first row of cooling passages. The radial cooling of the shield block avoids a complex by-pass at the back and opens up the possibility to use cast instead of forged steel. Slits in the shield block and in the first wall reduce the electromagnetic forces enough to allow the support of the modules on the vessel and the mechanical attachment of the first wall panels

  19. Adaptation of the HCPB DEMO TBM as breeding blanket for ITER : Neutronic and thermal analyses

    International Nuclear Information System (INIS)

    Aquaro, D.; Morellini, D.; Cerullo, N.

    2006-01-01

    Two breeding blanket are presently developed in Europe for the DEMO reactor: the first one, the Helium Cooled Lithium Lead (HCLL) uses a liquid breeder while the other , the Helium Cooled Pebble Bed (HCPB), uses a solid breeder in form of pebble bed. The modules of these blankets, called Test Blanket Modules (TBM) will be located in correspondence of the equatorial ports of ITER in order to be tested. ITER FEAT was designed with shielding blankets, therefore in the final stage of the experiment, in the foreseen tritium -deuterium operation phase, the tritium will be supplied to the reactor and not produced inside it. Since the production of tritium is of main importance for the feasibility of a nuclear fusion reactor, perhaps in the ITER final stage, the shielding blanket could be substituted by means of a breeding blanket. The geometry and composition of this breeding blanket would be, of course, similar to that of TBM which demonstrated to have the best performances. This paper illustrates a neutronic and thermal analysis of an hypothetical triziogen blanket for ITER FEAT made similar to a HCPB test module. The main aims of the performed analyses are to determine the Tritium Breeding Ratio (TBR) considering different solid breeders (Li 4 SiO 4 and Li 2 TiO 3 ) with different enrichment in 6 Li and different structural materials (a 9%CRWVTa reduced activation ferritic martensitic steel (EUROFER) or ceramic matrix composites like SiCf/SiC). The breeding blanket design is compared considering the highest value of TBR and the verification of the temperature constraints ( 550 o C for the steel, 950 o C for the breeder and 650 o C for the Beryllium). The neutronic analyses have been performed by means of MCNP-4C code and the thermal analyses using the MSC-MARC code. A TBR about equal 1 was obtained with a SiCf/SiC structural material and a Li 4 SiO 4 breeder. The performed analyses have to be considered preliminary and an academic exercise, nevertheless they could give

  20. Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

    Energy Technology Data Exchange (ETDEWEB)

    Pérez, Germán, E-mail: German.Perez@iter.org; Mitteau, Raphaël; Furmanek, Andreas; Martin, Alex; Raffray, René; Merola, Mario; Sabourin, Flavien

    2014-10-15

    Highlights: • Optimized water distribution in ITER blanket modules is presented. • All key challenging constraints are included. • The methodology and the successful result are presented. - Abstract: This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements. The key challenging constraints include to: be compatible with the overall water allocation (3140 kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70 °C at the outlet of each single BM; and ensure that water velocity is less than 7 m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented.

  1. Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

    International Nuclear Information System (INIS)

    Pérez, Germán; Mitteau, Raphaël; Furmanek, Andreas; Martin, Alex; Raffray, René; Merola, Mario; Sabourin, Flavien

    2014-01-01

    Highlights: • Optimized water distribution in ITER blanket modules is presented. • All key challenging constraints are included. • The methodology and the successful result are presented. - Abstract: This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements. The key challenging constraints include to: be compatible with the overall water allocation (3140 kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70 °C at the outlet of each single BM; and ensure that water velocity is less than 7 m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented

  2. Multi-purpose deployer for ITER in-vessel maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang-Hwan, E-mail: Chang-Hwan.CHOI@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Tesini, Alessandro; Subramanian, Rajendran [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Rolfe, Alan; Mills, Simon; Scott, Robin; Froud, Tim; Haist, Bernhard; McCarron, Eddie [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, OXON (United Kingdom)

    2015-10-15

    Highlights: • ITER RH system called as the multi-purpose deployer (MPD) is introduced. • The MPD performs dust and tritium inventory control, in-service inspection. • The MPD performs leak localization, in-vessel diagnostics maintenance. • The MPD has nine degrees of freedom with a payload capacity up to 2 tons. - Abstract: The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

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

    Science.gov (United States)

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

    1998-10-01

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

  5. The EC conceptual design proposal of a water-cooled convertible blanket for ITER

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Baraer, L.; Bielak, B.; Raepsaet, X.; Salavy, J.F.; Sedano, L.; Szczepanski, J.; Quintric-Bossy, J.; Severi, Y.

    1993-01-01

    For several years the EC laboratories have developed breeding blankets for DEMO. From this experience, it has been derived a proposal of tritium breeding blanket for the Extended Performance Phase (EPP) of ITER. The general basic ideas are the following: (i) the switch from the shielding blanket used during the BPP to the breeding blanket for the EPP should not require segments replacement ('convertible' blanket): (ii) its use should not have significant impact on the Basic Performance Phase (BPP); (iii) design and used materials should assure good safety standards and acceptable public perception; (iv) the blanket coolant should be compatible with the coolant required in the high heat-flux components (e.g. divertor, etc.; (v) the required R and D should fit with the ITER time schedule; (vi) the blanket should be able to withstand large power excursions and to accept long downtimes. The proposed design consists of a water-cooled liquid metal blanket, using the eutectic Pb-17Li during the EPP and a non-breeding Pb-alloy (Pb-18Mg or Pb-50Bi) during the BPP. Each segment is basically formed by a box containing the alloy, cooled by an array of poloidal hairpin-type cooling tubes and reinforced by toroidal and radial stiffeners. The coolant tubes are double-walled tubes allowing leak detections. The selected First Wall (FW) is a toroidally-drilled steel plate with brazed water-cooling U-tube. The structural material is austenitic stainless steel (316L(N)) which limits the maximum acceptable neutron fluence to about 1 MWa/m 2 . The advantages of using other structural materials requiring longer leadtimes, such as ferritic/martensitic steels, are also briefly discussed

  6. EU contribution to the procurement of the ITER blanket first wall

    International Nuclear Information System (INIS)

    Lorenzetto, Patrick; Banetta, Stefano; Bellin, Boris; Boireau, Bruno; Bucci, Philippe; Cicero, Tindaro; Conchon, Denis; Dellopoulos, Georges; Hardaker, Stephen; Marshall, Paul; Nogué, Patrice; Pérez, Marcos; Gutierrez, Leticia Ruiz; Samaniego, Fernando; Sherlock, Paul; Zacchia, Francesco

    2016-01-01

    Highlights: • Presentation of the blanket first wall design concept to be procured by Europe. • Presentation of the main outcome of the R&D programme with the resulting FW fabrication route. • Presentation of the ITER first wall pre-qualification programme with the results achieved so far. • Presentation of the on-going irradiation experiments. • Presentation of the EU procurement strategy. - Abstract: Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel. This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

  7. EU contribution to the procurement of the ITER blanket first wall

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzetto, Patrick, E-mail: Patrick.Lorenzetto@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Banetta, Stefano; Bellin, Boris [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Boireau, Bruno [AREVA NP, Centre Technique, 71200 Le Creusot (France); Bucci, Philippe [Atmostat, rue René Hamon 31, 94815 Villejuif Cedex (France); Cicero, Tindaro [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Conchon, Denis [Atmostat, rue René Hamon 31, 94815 Villejuif Cedex (France); Dellopoulos, Georges [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Hardaker, Stephen [Amec Foster Wheeler plc, Booths Park, Chelford Road, Knutsford WA16 8QZ (United Kingdom); Marshall, Paul [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Nogué, Patrice [AREVA NP, Centre Technique, 71200 Le Creusot (France); Pérez, Marcos [Leading Enterprises SL, Pasaje de La Agüera, 39409 San Felices de Buelna (Spain); Gutierrez, Leticia Ruiz [Iberdrola Ingeniería y Construcción S.A.U., Avenida Manoteras 20, 28050 Madrid (Spain); Samaniego, Fernando [Leading Enterprises SL, Pasaje de La Agüera, 39409 San Felices de Buelna (Spain); Sherlock, Paul [Amec Foster Wheeler plc, Booths Park, Chelford Road, Knutsford WA16 8QZ (United Kingdom); Zacchia, Francesco [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain)

    2016-11-01

    Highlights: • Presentation of the blanket first wall design concept to be procured by Europe. • Presentation of the main outcome of the R&D programme with the resulting FW fabrication route. • Presentation of the ITER first wall pre-qualification programme with the results achieved so far. • Presentation of the on-going irradiation experiments. • Presentation of the EU procurement strategy. - Abstract: Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel. This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

  8. The ITER EC H&CD Upper Launcher: Analysis of vertical Remote Handling applied to the BSM maintenance

    NARCIS (Netherlands)

    Grossetti, G.; Aiello, G.; Heemskerk, C.; Elzendoorn, B.; Geßner, R.; Koning, J.; Meier, A.; Ronden, D.; Späh, P.; Scherer, T.; Schreck, S.; Strauß, D.; Vaccaro, A.

    2013-01-01

    This paper deals with Remote Handling activities foreseen on the Blanket Shield Module, the plasma facing component of the ITER Electron Cyclotron Heating and Current Drive Upper Launcher. The maintenance configuration considered here is the Vertical Remote Handling, meaning gravity acting along the

  9. Current status on detail design and fabrication techniques development of ITER blanket shield block in Korea

    International Nuclear Information System (INIS)

    Kim, Duck Hoi; Cho, Seungyon; Ahn, Mu-Young; Lee, Eun-Seok; Jung, Ki Jung

    2007-01-01

    The allocation of components and systems to be delivered to ITER on an in-kind basis, was agreed between the ITER Parties. Among parties, Korea agreed to procure inboard blanket modules 1, 2 and 6, which consists of FW and shield block. Regarding shield block the detail design and Fabrication techniques development have been undertaken in Korea. Especially manufacturing feasibility study on shield block had been performed and some technical issues for the fabrication were selected. Based on these results, fabrication techniques using EB welding are being developed. Meanwhile, the detail design of inboard standard module has been carried out. The optimization of flow driver design to improve the cooling performance was executed. And, thermo-hydraulic analysis on half block of inboard standard module was performed. In this study, current status and some results from Fabrication techniques development on ITER blanket shield block are described. The detail design activity and results on shield block are also introduced herein. (orig.)

  10. Detailed mechanical design and manufacturing study for the ITER reference breeding blanket

    International Nuclear Information System (INIS)

    Zacchia, F.; Daenner, W.; Stefanis, L. de; Ferrari, M.; Gerber, A.; Mustoe, J.

    1998-01-01

    This papers relates on the detailed mechanical design, manufacturing feasibility and assembly analysis of a water-cooled solid breeding blanket concept, selected as the ITER reference design. This breeding blanket design is characterised by: i) pressurised water flowing inside flat steel panels for cooling of the internals; each panel is welded along its contour onto the first wall structure and to the rear shield plate after closure of the module (last assembly step). ii) Beryllium (neutronic multiplier) in the form of micro-spheres filling the volume between parallel flat coolant panels. iii) Breeder pebbles enclosed in rods, which form bundles and are themselves embedded inside the Beryllium micro-spheres. (authors)

  11. Nuclear, thermo-mechanical and tritium release analysis of ITER breeding blanket

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Miki, Nobuharu; Akiba, Masato

    2003-06-01

    The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket. With respect to the neutronics analysis, the detailed analyses of the distribution of the nuclear heating rate and TBR have been performed in 2D model using MCNP to clarify the input data for the tritium inventory and release rate analyses and thermo-mechanical analyses. With respect to the tritium inventory and release behavior analysis, the parametric analyses for selection of purge gas flow rate were carried out from the view point of pressure drop and the tritium inventory/release performance for Li 2 TiO 3 breeder. The analysis result concluded that purge gas flow rate can be set to conventional flow rate setting (88 l/min per module) to 1/10 of that to save the purge gas flow and minimize the size of purge gas pipe. However, it is necessary to note that more tritium is transformed to HTO (chemical form of water) in case of Li 2 TiO 3 compared to other breeder materials. With respect to the thermo-mechanical analyses of the pebble bed blanket structure, the analyses have been performed by ABAQUS with 2D model derived from one of eight facets of a blanket module, based on the reference design. Analyses were performed to identify the temperature distribution incorporating the pebble bed mechanical simulation and influence of mechanical behavior to the thermal behavior. The result showed that the maximum temperature in the breeding material was 617degC in the first row of breeding rods and the minimum temperature was 328

  12. Status of ITER blanket attachment design and related R and D

    International Nuclear Information System (INIS)

    Sadakov, S.; Khomiakov, S.; Calcagno, B.; Chappuis, Ph.; Dellopoulos, G.; Kolganov, V.; Merola, M.; Poddubnyi, I.; Raffray, R.; Raharijaona, J.J.; Ulrickson, M.; Zhmakin, A.

    2013-01-01

    Highlights: • ITER blanket attachment system went through a significant design upgrade and become basically compliant with specified design loads and required cyclic lifetime. • Upgrade of flexible supports allowed the doubling of cross sections of central bolts. Ceramic coatings were relocated to much larger areas on conical pairs screwed into shield blocks. • Key pads were relocated from keys of vacuum vessel into keyways of shield blocks and reshaped to enlarge areas of lateral interfaces with ceramic electro-insulating coatings. • Ceramic coatings are hidden between pads and enclosures in keyways with a purpose to minimize their wear rate, which depends on peak friction stress and cyclic sliding path. • Ceramic coatings to be verified by experiment, with several R and D aimed to collect statistically sufficient data on their reliability and durability in ITER relevant cyclic loading conditions. -- Abstract: Main function of the ITER blanket system [1–3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R and D is also briefly described

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-01

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

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

    Science.gov (United States)

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

    2000-12-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  20. DEMO maintenance scenarios: scheme for time estimations and preliminary estimates for blankets arranged in multi-module-segments

    International Nuclear Information System (INIS)

    Nagy, D.

    2007-01-01

    Previous conceptual studies made clear that the ITER blanket concept and segmentation is not suitable for the environment of a potential fusion power plant (DEMO). One promising concept to be used instead is the so-called Multi-Module-Segment (MMS) concept. Each MMS consists of a number of blankets arranged on a strong back plate thus forming ''banana'' shaped in-board (IB) and out-board (OB) segments. With respect to port size, weight, or other limiting aspects the IB and OB MMS are segmented in toroidal direction. The number of segments to be replaced would be below 100. For this segmentation concept a new maintenance scenario had to be worked out. The aim of this paper is to present a promising MMS maintenance scenario, a flexible scheme for time estimations under varying boundary conditions and preliminary time estimates. According to the proposed scenario two upper, vertical arranged maintenance ports have to be opened for blanket maintenance on opposite sides of the tokamak. Both ports are central to a 180 degree sector and the MMS are removed and inserted through both ports. In-vessel machines are operating to transport the elements in toroidal direction and also to insert and attach the MMS to the shield. Outside the vessel the elements have to be transported between the tokamak and the hot cell to be refurbished. Calculating the maintenance time for such a scenario is rather challenging due to the numerous parallel processes involved. For this reason a flexible, multi-level calculation scheme has been developed in which the operations are organized into three levels: At the lowest level the basic maintenance steps are determined. These are organized into maintenance sequences that take into account parallelisms in the system. Several maintenance sequences constitute the maintenance phases which correspond to a certain logistics scenario. By adding the required times of the maintenance phases the total maintenance time is obtained. The paper presents

  1. Concept for a vertical maintenance remote handling system for multi module blanket segments in DEMO

    International Nuclear Information System (INIS)

    Coleman, M.; Sykes, N.; Cooper, D.; Iglesias, D.; Bastow, R.; Loving, A.; Harman, J.

    2014-01-01

    Highlights: •A conceptual architectural model for a vertical maintenance DEMO is presented. •Novel concepts for a set of DEMO remote handling equipment are put forward. •Remote maintenance of a multi module segment blanket is found to be feasible. •The criticality of space in the vertical port is highlighted. -- Abstract: The anticipated high neutron flux, and the consequent damage to plasma-facing components in DEMO, results in the need to regularly replace the tritium breeding and radiation shielding blanket. The current European multi module segment (MMS) blanket concept favours a less invasive small port entry maintenance system over large sector transport concepts, because of the reduced impact on other tokamak systems – particularly the magnetic coils. This paper presents a novel conceptual remote maintenance strategy for a Vertical Maintenance Scheme DEMO, incorporating substantiated designs for an in-vessel mover, to detach and attach the blanket segments, and cask-housed vertical maintenance devices to open and close access ports, cut and join service connections, and extract blanket segments from the vessel. In addition, a conceptual architectural model for DEMO was generated to capture functional and spatial interfaces between the remote maintenance equipment and other systems. Areas of further study are identified in order to comprehensively establish the feasibility of the proposed maintenance system

  2. Engineering challenges and development of the ITER Blanket System and Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, Alphonse Rene; Chappuis, Philippe; Hirai, Takeshi; Gicquel, Stefan

    2015-10-15

    The ITER Blanket System and the Divertor are the main components which directly face the plasma. Being the first physical barrier to the plasma, they have very demanding design requirements, which include accommodating: (1) surface heat flux and neutronic volumetric heating, (2) electromagnetic loads, (3) nuclear shielding function, (4) capability of being assembled and remote-handled, (5) interfaces with other in-vessel components, and (6) high heat flux technologies and complex welded structures in the design. The main functions of the Blanket System have been substantially expanded and it has now also to provide limiting surfaces that define the plasma boundary during startup and shutdown. As regards the Divertor, the ITER Council decided in November 2013 to start the ITER operation with a full-tungsten armour in order to minimize costs and already gain operational experience with tungsten during the non-active phase of the machine. This paper gives an overview of the design and technology qualification of the Blanket System and the Divertor.

  3. Remote maintenance challenges presented in the ITER engineering design

    International Nuclear Information System (INIS)

    Burgess, T.W.; Herndon, J.N.; Schrock, S.L.; Lousteau, D.C.

    1995-01-01

    Leading fusion energy research institutions are currently engaged in the Engineering Design Activity for the International Thermonuclear Experimental Reactor (ITER). A tokamak reactor design is evolving which emphasizes high system performance in a minimum overall reactor and building size. The resulting high component density dictates careful attention to ITER remote maintenance considerations in the development of the configuration. The complexity and scale of ITER remote maintenance tasks are well beyond the scope of today's experience and technology. This paper discusses the remote maintenance philosophy, describes the basic configuration as it relates to maintenance, and describes the basic procedures and equipment required. Key enabling technology research and development needs are also addressed

  4. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    International Nuclear Information System (INIS)

    Hong, Bong Guen; In, S. R.; Bae, Y. D.

    2006-02-01

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  5. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; In, S. R.; Bae, Y. D. (and others)

    2006-02-15

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  6. Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

    International Nuclear Information System (INIS)

    Tanigawa, H.; Hirose, T.; Shiba, K.; Kasada, R.; Wakai, E.; Serizawa, H.; Kawahito, Y.; Jitsukawa, S.; Kimura, A.; Kohno, Y.; Kohyama, A.; Katayama, S.; Mori, H.; Nishimoto, K.; Klueh, R.L.; Sokolov, M.A.; Stoller, R.E.; Zinkle, S.J.

    2008-01-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed

  7. Assistance tools for generic definition of ITER maintenance tasks and scenarios in advanced supervisory control systems

    International Nuclear Information System (INIS)

    Zieba, Stéphane; Russotto, François-Xavier; Da Silva Simoes, Max; Measson, Yvan

    2013-01-01

    Highlights: ► Improve supervisory control systems for ITER in-vessel and hot cell maintenance. ► Optimize remote handling operations effectiveness, reliability and safety. ► Provide a generic description of the maintenance tasks and scenarios. ► Development of context-based assistances for operators and supervisor. ► Improvement of operator's situation awareness. -- Abstract: This paper concerns the improvement of supervisory control systems in the context of remote handling for the maintenance tasks in ITER. This work aims at providing a single formalism and tools to define in a generic way the ITER maintenance tasks and scenarios for in-vessel and hot cell operations. A three-layered approach is proposed to model these tasks and scenarios. Physical actions are defined for the scene elements. From these physical actions, behaviours are defined to represent high-level functionalities. Finally, interaction modes define the way that behaviours are achieved in terms of human–machine interactions. Case study concerning the blanket maintenance procedure is discussed concerning the contributions of the descriptive model and the context-based assistances to the activities of supervisory control

  8. Conceptual design and testing strategy of a dual functional lithium-lead test blanket module in ITER and EAST

    International Nuclear Information System (INIS)

    Wu, Y.

    2007-01-01

    A dual functional lithium-lead (DFLL) test blanket module (TBM) concept has been proposed for testing in the International Thermonuclear Experimental Reactor (ITER) and the Experimental Advanced Superconducting Tokamak (EAST) in China to demonstrate the technologies of the liquid lithium-lead breeder blankets with emphasis on the balance between the risks and the potential attractiveness of blanket technology development. The design of DFLL-TBM concept has the flexibility of testing both the helium-cooled quasi-static lithium-lead (SLL) blanket concept and the He/PbLi dual-cooled lithium-lead (DLL) blanket concept. This paper presents an effective testing strategy proposed to achieve the testing target of SLL and DLL DEMO blankets relevant conditions, which includes three parts: materials R and D and small-scale out-of-pile mockups testing in loops, middle-scale TBMs pre-testing in EAST and full-scale consecutive TBMs testing corresponding to different operation phases of ITER during the first 10 years. The design of the DFLL-TBM concept and the testing strategy ability to test TBMs for both blanket concepts in sequence and or in parallel for both ITER and EAST are discussed

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. 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)

  11. ANL ITER high-heat-flux blanket-module heat transfer experiments

    International Nuclear Information System (INIS)

    Kasza, K.E.

    1992-02-01

    An Argonne National Laboratory facility for conducting tests on multilayered slab models of fusion blanket designs is being developed; some of its features are described. This facility will allow testing under prototypic high heat fluxes, high temperatures, thermal gradients, and variable mechanical loadings in a helium gas environment. Steady and transient heat flux tests are possible. Electrical heating by a two-sided, thin stainless steel (SS) plate electrical resistance heater and SS water-cooled cold panels placed symmetrically on both sides of the heater allow achievement of global one-dimensional heat transfer across blanket specimen layers sandwiched between the hot and cold plates. The heat transfer characteristics at interfaces, as well as macroscale and microscale thermomechanical interactions between layers, can be studied in support of the ITER engineering design effort. The engineering design of the test apparatus has shown that it is important to use multidimensional thermomechanical analysis of sandwich-type composites to adequately analyze heat transfer. This fact will also be true for the engineering design of ITER

  12. Summary report for ITER task - T68: MHD facility preparation for Li/V blanket option

    International Nuclear Information System (INIS)

    Reed, C.B.; Haglund, R.C.; Miller, M.E.

    1995-08-01

    A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the question of insulator coatings. Design calculations show that an electrically insulating layer is necessary to maintain an acceptably low MHD pressure drop. To enable experimental investigations of the MHD performance of candidate insulator materials and the technology for putting them in place, the room-temperature ALEX (Argonne's Liquid Metal EXperiment) NaK facility was upgraded to a 300 degrees C lithium system. The objective of this upgrade was to modify the existing facility to the minimum extent necessary, consistent with providing a safe, flexible, and easy to operate MHD test facility which uses lithium at ITER-relevant temperatures, Hartmann numbers, and interaction parameters. The facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups. The system design description for this lithium upgrade of the ALEX facility is given in this document

  13. ITER Blanket First Wall (WBS 1.6{sub 1}A)

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-15

    International Thermonuclear Experimental Reactor (ITER) project is the international collaboration one for the commercialization of nuclear fusion energy through the technical and engineering verification. In ITER project, we plan to procure the blanket systems which has the risk of technology and cost when it is newly developed. We are developing the manufacturing process and joining technology for the ITER blanket to complete the procurement with qualified blanket system. To evaluate the soundness of manufacturing process, specimen and mock-up tests are being prepared. Finally, we can obtain the key technology of nuclear fusion reactor especially on the blanket design, joining and manufacturing technology through the present project and these technologies will help the construction of Korea fusion DEMO reactor and the development of commercial nuclear fusion reactor in Korea. In 1st year, through the fabrication of the Cu/SS and Be/Cu joint specimen, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The optimized HIP conditions (1050 .deg. C, 150 MPa, 2 hr for Cu/SS and 580 - 620 .deg. C, 100-150 MPa, 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint and NDT such as UT (10 MHz, 0.25 inch D, flat type) and ECT. Several mock-ups were fabricated for confirming the joint integrity and NDT. specimens fabricated with these mock-ups were used in mechanical tests including microstructure observation. The mock-ups were used in the HHF test after the developed NDT. In 2nd year, PHHT of Cu was investigated in order to recover its mechanical properties, and the pre-qualification mock-up were fabricated against the Qualification Program and sent to RF for HHF testing in TSEFEY. FW fabrication and joining procedure were documented in the form of the TSD. Qualification mock

  14. Current Status on the Korean Test Blanket Module Development for testing in the ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Jung, Ki Sok

    2010-01-01

    Korea has proposed and designed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER). Ferrite Martensitic (FM) steel is used as the structural material and helium (He) is used as a coolant to cool the first wall (FW) and breeding zone. Liquid lithium (Li) is circulated for a tritium breeding, not for a cooling purpose. Main purpose for developing the TBM is to develop the design technology for DEMO and fusion reactor and it should be proved through the experiment in the ITER with TBM. Therefore, we have developed the design scheme and related codes including the safety analysis for obtain the license to be tested in the ITER. In order to develop and install at the ITER, several technologies were developed in parallel; fabrication, breeder, He cooling, tritium extraction and so on. Figure 1 shows the overall TBM development scheme. In Korea, official strategy for developing the TBM is to participate to other parties' concept such as US and EU ones, in which PbLi (lead lithium eutectic), He, and FM steel were used for liquid breeder, coolant, and structural material, respectively

  15. Activation and afterheat analyses for the HCPB test blanket module in ITER

    International Nuclear Information System (INIS)

    Pereslavtsev, P.; Fischer, U.

    2008-01-01

    To provide a sound data basis for the safety analyses of the HCPB TBM system in ITER, the afterheat and activity inventories were assessed making use of a code system that allows performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. A suitable MCNP model of a 20 deg. ITER torus sector with an integrated TBM of the HCPB PI (plant integration) type in the horizontal test blanket port was developed and adapted to the requirements for coupled 3D neutron transport and activation calculations. Two different irradiation scenarios were considered in the coupled 3D neutron transport and activation calculations. The first one is representative for the TBM irradiation in ITER with a total of 9000 neutron pulses over a 3 (calendar) years period. The second (conservative) irradiation scenario assumes an extended irradiation time over the full anticipated lifetime of ITER. The radioactivity inventories, the afterheat and the contact gamma dose were calculated as function of the decay time. Data were processed for the total activity, afterheat and contact dose rates of the TBM, its constituting components and materials

  16. ITER EDA newsletter. V. 9, no. 9

    International Nuclear Information System (INIS)

    2000-09-01

    This ITER EDA Newsletter contains the following 5 contributions: CSMC and CSIC charging tests successfully completed; The ITER divertor cassette project meeting; Blanket R and D and design task meeting; IAEA technical committee meeting on fusion safety; ITER L-6 large project ''blanket remote handling and maintenance''

  17. Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

    Science.gov (United States)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

    2006-02-01

    This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li2TiO3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 °C followed by normalizing it at 930 °C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li2TiO3, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li2TiO3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li2TiO3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

  18. Key achievements in elementary R and D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

    International Nuclear Information System (INIS)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

    2006-01-01

    This paper presents the significant progress made in the research and development (R and D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li 2 TiO 3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 0 C followed by normalizing it at 930 0 C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R and D on the breeder material, Li 2 TiO 3 , the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li 2 TiO 3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li 2 TiO 3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation

  19. Key achievements in elementary R and Ds on water-cooled solid breeder blanket for ITER Test Blanket Module in JAERI

    International Nuclear Information System (INIS)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Tanigawa, H.; Tobita, K.; Akiba, M.; Hayashi, K.; Ochiai, K.; Nishitani, T.

    2005-01-01

    This paper presents significant progress in research and development (R and D) of key elementary technologies on the water-cooled solid breeder blanket for the ITER test blanket modules (TBMs) in JAERI. Development of module fabrication technology, bonding technology of armors, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup, and tritium release behavior from Li 2 TiO 3 pebble bed under neutron pulsed operation condition are summarized. By the improvement of heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H, can be obtained by homogenizing it at 1150 deg C followed by normalizing at 930 deg C after the Hot Isostatic Pressing (HIP) process. Moreover, a promising bonding process for a tungsten armor and an F82H structural material was developed by using a solid state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it was found that the thermal fatigue lifetime of F82H can be predicted by using Manson-Coffin's law. As for R and Ds on a breeder material, Li 2 TiO 3 , effective thermal conductivity of Li 2 TiO 3 pebble was measured under compressive force simulating the ITER TBM environment. The increase in the effective thermal conductivity of the pebble bed was about 2.5 % at the compressive strain of 0.9 % at 400 deg C. Neutronic performance of the blanket module mockup has been carried out by the 14 MeV neutron irradiation. It was confirmed that the measured tritium production rate agreed with the calculated values within about 10% difference. Also, tritium release from a Li 2 TiO 3 pebble bed was measured under pulsed neutron irradiation conditions simulating the ITER operation. (author)

  20. Shutdown dose rate analysis of European test blanket modules shields in ITER Equatorial Port #16

    Energy Technology Data Exchange (ETDEWEB)

    Juárez, Rafael, E-mail: rjuarez@ind.uned.es [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain); Sauvan, Patrick; Perez, Lucia [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain); Panayotov, Dobromir; Vallory, Joelle; Zmitko, Milan; Poitevin, Yves [Fusion for Energy (F4E), Torres Diagonal Litoral B3, Josep Pla 2, Barcelona 08019 (Spain); Sanz, Javier [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain)

    2016-11-01

    Highlights: • Nuclear analysis for European TBMs and shields, in ITER Equatorial Port #16, has been conducted in support of the ‘Concept Design Review’ from ITER. • The objective of the work is the characterization of the Shutdown Dose Rates at Equatorial Port #16 interspace. • The role played by the TBM and TBM shields, the equatorial port gaps and the vacuum vessel permeation, in terms of neutron flux transmission is assessed. • The role played by the TBM, TBM shields, Port Plug Frame, Pipe Forest and the machine in terms of activation is also investigated. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). An essential element of the Conceptual Design Review (CDR) of these TBSs is the demonstration of capability of Test Blanket Modules (TBM) and their shields to fulfil their function and comply with the design requirements. One of the TBM shields highly relevant design aspects is the project target for shutdown dose rates (SDDR) in the interspace. We investigated two functions of the TBMs and TBM shields—the neutron flux attenuation along the shields, and the reduction of the activation of the components contributing to SDDR. It is shown that TBMs and TBM shields reduce significantly the neutron flux in the port plug (PP). In terms of neutron flux attenuation, the TBM shield provides sufficient neutron flux reduction, being responsible for 5 × 10{sup 6} n/cm{sup 2} s at port interspace, while the EPP gaps and BSM gaps are responsible for 5 × 10{sup 7} n/cm{sup 2} s each. When considering closed upper, lower and lateral neighbour equatorial ports (thus, excluding the cross-talk between ports), a SDDR of 121 μSv/h averaged near the port closure flange was obtained, out of which, only 4 μSv/h are due to the activation of TBMs and TBM shields. Maximum SDDR in the range

  1. Experimental results and validation of a method to reconstruct forces on the ITER test blanket modules

    International Nuclear Information System (INIS)

    Zeile, Christian; Maione, Ivan A.

    2015-01-01

    Highlights: • An in operation force measurement system for the ITER EU HCPB TBM has been developed. • The force reconstruction methods are based on strain measurements on the attachment system. • An experimental setup and a corresponding mock-up have been built. • A set of test cases representing ITER relevant excitations has been used for validation. • The influence of modeling errors on the force reconstruction has been investigated. - Abstract: In order to reconstruct forces on the test blanket modules in ITER, two force reconstruction methods, the augmented Kalman filter and a model predictive controller, have been selected and developed to estimate the forces based on strain measurements on the attachment system. A dedicated experimental setup with a corresponding mock-up has been designed and built to validate these methods. A set of test cases has been defined to represent possible excitation of the system. It has been shown that the errors in the estimated forces mainly depend on the accuracy of the identified model used by the algorithms. Furthermore, it has been found that a minimum of 10 strain gauges is necessary to allow for a low error in the reconstructed forces.

  2. Maintenance schemes for the ITER neutral beam test facility

    International Nuclear Information System (INIS)

    Zaccaria, P.; Dal Bello, S.; Marcuzzi, D.; Masiello, A.; Coniglio, A.; Antoni, V.; Cordier, J.J.; Hemsworth, R.; Jones, T.; Di Pietro, E.; Mondino, P.L.

    2004-01-01

    The ITER neutral beam test facility (NBTF) is planned to be built, after the approval of the ITER construction and the choice of the ITER site, with the agreement of the ITER International Team and of the JA and RF participant teams. The key purpose is to progressively increase the performance of the first ITER injector and to demonstrate its reliability at the maximum operation parameters: power delivered to the plasma 16.5 MW, beam energy 1 MeV, accelerated D - ion current 40 A, pulse length 3600 s. Several interventions for possible modifications and for maintenance are expected during the early operation of the ITER injector in order to optimize the beam generation, aiming and steering. The maintenance scheme and the related design solutions are therefore a very important aspect to be considered for the NBTF design. The paper describes consistently the many interrelated aspects of the design, such as the optimisation of the vessel and cryopump geometry, in order to get a better maintenance flexibility, an easier man access and a larger access for diagnostic and monitoring. (authors)

  3. Dynamic test of the ITER blanket key and ceramic insulated pad

    International Nuclear Information System (INIS)

    Khomyakov, S.; Sysoev, G.; Strebkov, Yu.; Kucherov, A.; Ioki, K.

    2010-01-01

    The dynamic testing of the blanket module's key integrated into ITER vacuum vessel portion has been performed in 2008 to investigate its capability to react the electro-magnetic (EM) loads. The preliminary analysis showed the large dynamic amplification factor (DAF) of the reactions because of technological gaps between the blanket module and key. Shock load may yield the bronze pads, which protect the blanket electrical insulation from damage. However the dynamic analysis of such particularly non-linear system needs an experimental ground and confirmation. Toward this end, as well as demonstration of the key reliability, the special test facility has been made, and the full-scale mock-up of the inboard intermodular key was tested. So as not to scale non-linear dynamic parameters, 1-ton mass was built on the single flexible support. The key was welded in a 60-mm thick steel plate modeled with a fragment of the VV. The different gaps were set in between the bronze pad of the key and the mass shock worker. This system (supplemented with some additional constraints) has natural oscillations like as the 4-ton module built on four flexible supports. Thus the most critical radial torque might be modeled with a straight force. The objectives of the test were as follows: dynamic response, DAF and damping factor determination; measurement of the strain oscillations in the key's base and in the weld seam; comparison of the measured data with computation results. The paper will present the analytical grounds of the testing conditions, test facility description, analytical adaptation of the facility, experimental results, its comparison with analysis and discussion, and guidelines for the next experimental phase.

  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. Experimental and numerical study of the pressure drop for ITER blanket shield block

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Min-Su; Kim, Sawoong; Jung, Hun-Chea; Shim, Hee-Jin; Ahn, Hee-Jae

    2016-11-01

    Highlights: • The results of the experiment and the numerical analysis are compared. • The numerical analysis results are lower than the experimental results. • The margin of the pressure drop is suggested. - Abstract: The blanket shield block (SB) is located inside the ITER vacuum chamber, and the main function is to provide the thermal and nuclear shielding to the vacuum vessel and external components. The SB is foreseen to undergo a significant heat load which is a body load throughout the whole thickness of the SB under normal operation conditions. Therefore, the cooling configuration in SB should be designed very carefully based on the various experiences. The pressure drop in the cooling design is one of the most important factors to balance a water distribution of overall blanket cooling system. In order to verify the pressure drop characteristic and validate the design methodology of SB, experiment and numerical analysis are performed and compared their results. These results would be a benchmarking of the numerical results with experimental results to assess the gap between calculations and experiments.

  6. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures

    International Nuclear Information System (INIS)

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

    1994-01-01

    Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250 degrees C. These specimens have been tested over a temperature range from 20 to 250 degrees C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250 degrees C is more damaging than at 90 degrees C, causing larger decreases in the fracture toughness. Ferritic-martensitic steels are embrittled by the irradiation, and show the lowest toughness at room temperature

  7. Activation analysis and waste management of China ITER helium cooled solid breeder test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Han, J.R., E-mail: hanjingru@163.co [North China Electric Power University, School of Nuclear Science and Engineering, Zhu-Xin-Zhuang, De-Wai, Beijing 102206 (China); Chen, Y.X.; Han, R. [North China Electric Power University, School of Nuclear Science and Engineering, Zhu-Xin-Zhuang, De-Wai, Beijing 102206 (China); Feng, K.M. [Southwestern Institute of Physics, P.O.Box 432, Chengdu 610041 (China); Forrest, R.A. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom)

    2010-08-15

    Activation characteristics have been assessed for the ITER China helium cooled solid breeder (CH-HCSB) 3 x 6 test blanket module (TBM). Taking a representative irradiation scenario, the activation calculations were performed by FISPACT code. Neutron fluxes distributions in the TBM were provided by a preceding MCNP calculation. These fluxes were passed to FISPACT for the activation calculation. The main activation parameters of the HCSB-TBM were calculated and discussed, such as activity, afterheat and contact dose rate. Meanwhile, the dominant radioactivity nuclides and reaction channel pathways have been identified. According to the Safety and Environmental Assessment of Fusion Power (SEAFP) waste management strategy, the activated materials can be re-used following the remote handling recycling options. The results will provide useful indications for further optimization design and waste management of the TBM.

  8. ITER structural design criteria and their extension to advanced reactor blankets

    International Nuclear Information System (INIS)

    Majumdar, S.; Kalinin, G.

    2000-01-01

    Applications of the recent ITER structural design criteria (ISDC) are illustrated by two components. First, the low-temperature-design rules are applied to copper alloys that are particularly prone to irradiation embrittlement at relatively low fluences at certain temperatures. Allowable stresses are derived and the impact of the embrittlement on allowable surface heat flux of a simple first-wall/limiter design is demonstrated. Next, the high-temperature-design rules of ISDC are applied to evaporation of lithium and vapor extraction (EVOLVE), a blanket design concept currently being investigated under the US Advanced Power Extraction (APEX) program. A single tungsten first-wall tube is considered for thermal and stress analyses by finite-element method

  9. Applying principles of Design For Assembly to ITER maintenance operations

    International Nuclear Information System (INIS)

    Heemskerk, Cock; de Baar, Marco; Elzendoorn, Ben; Koning, Jarich; Verhoeven, Toon; Vreede, Fred de

    2009-01-01

    In ITER, maintenance operations in the vessel and in the Hot Cell will be largely done by Remote Handling (RH). Remotely performed maintenance actions tend to be more time-costly than actions performed by direct human access. With a human operator in the control loop and adequate situational feedback, a two-armed master slave manipulator system can mimic direct access with dexterous manipulation, tactile feedback and vision. But even then, turnaround times are still very high. Adapting the design for simplified maintenance operations can yield significant time savings. One of the methods known to produce a simpler, more robust design, which is also better suited for handling with robots, is Design For Assembly (DFA). This paper discusses whether and how the principles of DFA can be applied to simplify maintenance operations for ITER. While DFA is normally used with series-production and ITER is a unique product, it is possible to apply the principles of DFA to ITER maintenance operations. Furthermore, DFA's principles can be applied at different abstraction levels. Combining principles of DFA with Virtual Reality leads to new insights and provides additional value.

  10. Japanese contribution to the design of primary module of shielding blanket in ITER-FEAT

    International Nuclear Information System (INIS)

    Kuroda, Toshimasa; Hatano, Toshihisa; Miki, Nobuharu; Hiroki, Seiji; Enoeda, Mikio; Ohmori, Junji; Akiba, Masato

    2003-02-01

    Japanese contributions to the design activity on the shielding blanket module consisting of the separable first wall and the shield block for ITER-FEAT are compiled. Temperature and stress distributions in the first wall and the shield block are analyzed and evaluated with 2-D and 3-D models for steady state and also for transient condition according to plasma ramp-up and ramp-down. While temperatures and stresses in the first wall satisfy their allowable values, those in a front part of the shield block exceed the allowable guideline. Based on this result, design improvements are suggested. Coolant flow and pressure distributions along the complicated coolant channel in the shield block are preliminary analyzed. Though heat removal is satisfactory in all coolant channels, back flows due to choking in coolant collectors are found. Design improvements to avoid the choking are suggested. Electromagnetic forces acting on blanket modules are analyzed with detailed 3-D models of solid elements for different disruption scenarios. The maximum moment around radial axis is 1.36 MNm on module no.5 under fast upward VDE, and the maximum moment around vertical axis is 1.47 MNm on module no.1 under fast downward VDE. The supporting beam of the first wall with welded attachment to the shield block is designed. Required welding thickness and support conditions to withstand electromagnetic forces are estimated. Strength of the shield block at the region mating the flexible cartridge is also estimated. Though the shield block surface attached by the flexible cartridge shows sufficient strength, the internal thread mating the Inconel bolt would need more length. In addition, water-to-water leak detection system in case main supply/return manifolds are located within the vacuum vessel is designed. By using Kr as the tracer material, the possibility of water-to-water leak detection and the concept of the detection system are shown. The design of the shielding blanket of ITER-FEAT has

  11. European Helium Cooled Pebble Bed (HCPB) test blanket. ITER design description document. Status 1.12.1996

    International Nuclear Information System (INIS)

    Albrecht, H.; Boccaccini, L.V.; Dalle Donne, M.; Fischer, U.; Gordeev, S.; Hutter, E.; Kleefeldt, K.; Norajitra, P.; Reimann, G.; Ruatto, P.; Schleisiek, K.; Schnauder, H.

    1997-04-01

    The Helium Cooled Pebble Bed (HCPB) blanket is based on the use of separate small lithium orthosilicate and beryllium pebble beds placed between radial toroidal cooling plates. The cooling is provided by helium at 8 MPa. The tritium produced in the pebble beds is purged by the flow of helium at 0.1 MPa. The structural material is martensitic steel. It is foreseen, after an extended R and D work, to test in ITER a blanket module based on the HCPB design, which is one of the two European proposals for the ITER Test Blanket Programme. To facilitate the handling operation the Blanket Test Module (BTM) is bolted to a surrounding water cooled frame fixed to the ITER shield blanket back plate. For the design of the test module, three-dimensional Monte Carlo neutronic calculations and thermohydraulic and stress analyses for the operation during the Basic Performance Phase (BPP) and during the Extended Performance Phase (EPP) of ITER have been performed. The behaviour of the test module during LOCA and LOFA has been investigated. Conceptual designs of the required ancillary loops have been performed. The present report is the updated version of the Design Description Document (DDD) for the HCPB Test Module. It has been written in accordance with a scheme given by the ITER Joint Central Team (JCT) and accounts for the comments made by the JCT to the previous version of this report. This work has been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhne and it is supported by the European Union within the European Fusion Technology Program. (orig.) [de

  12. Development of pipe welding, cutting and inspection tools for the ITER blanket

    International Nuclear Information System (INIS)

    Oka, Kiyoshi; Ito, Akira; Taguchi, Kou; Takiguchi, Yuji; Takahashi, Hiroyuki; Tada, Eisuke

    1999-07-01

    In D-T burning reactors such as International Thermonuclear Experimental Reactor (ITER), an internal access welding/cutting of blanket cooling pipe with bend sections is inevitably required because of spatial constraint due to nuclear shield and available port opening space. For this purpose, internal access pipe welding/cutting/inspection tools for manifolds and branch pipes are being developed according to the agreement of the ITER R and D task (T329). A design concept of welding/cutting processing head with a flexible optical fiber has been developed and the basic feasibility studies on welding, cutting and rewelding are performed using stainless steel plate (SS316L). In the same way, a design concept of inspection head with a non-destructive inspection probe (including a leak-testing probe) has been developed and the basic characteristic tests are performed using welded stainless steel pipes. In this report, the details of welding/cutting/inspection heads for manifolds and branch pipes are described, together with the basic experiment results relating to the welding/cutting and inspection. In addition, details of a composite type optical fiber, which can transmit both the high-power YAG laser and visible rays, is described. (author)

  13. Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall

    International Nuclear Information System (INIS)

    Kim, Byoung-Yoon; Ahn, Hee-Jae

    2011-01-01

    The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.

  14. Activation analysis of Chinese ITER helium cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Han Jingru; Chen Yixue; Ma Xubo; Wang Shouhai; Forrest, R.A.

    2009-01-01

    Based on the Chinese ITER helium cooled solid breeder(CH-HCSB) test blanket module (TBM) of the 3 x 6 sub-modules options, the activation characteristics of the TBM were calculated. Three-dimensional neutronic calculations were performed using the Monte-Carlo code MCNP and the nuclear data library FENDL/2. Furthermore, the activation calculations of HCSB-TBM were carried out with the European activation system EASY-2007. At shutdown the total activity is 1.29 x 10 16 Bq, and the total afterheat is 2.46 kW. They are both dominated by the Eurofer steel. The activity and afterheat are both in the safe range of TBM design, and will not have a great impact on the environment. Meanwhile,on basis of the calculated contact dose rate, the activated materials can be re-used following the remote handling recycling options. The activation results demonstrate that the current HCSB-TBM design can satisfy the ITER safety design requirements from the activation point of view. (authors)

  15. Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket

    International Nuclear Information System (INIS)

    Spannagel, G.; Gierszewski, P.

    1991-01-01

    At the Karlsruhe Nuclear Research Center (KfK) a flexible tool is being developed to simulate the dynamics of tritium inventories. This tool can be applied to any tritium handling system, especially to the fuel cycle components of future nuclear fusion devices. This instrument of simulation will be validated in equipment to be operated at the Karlsruhe Tritium Laboratory. In this study tritium inventories in a NET/ITER type fuel cycle involving a lithium salt solution blanket are investigated. The salt solution blanket serves as an example because it offers technological properties which are attractive in modeling the process; the example does not impair the general validity of the tool. Usually, the operation strategy of complex structures will deteriorate due to failures of the subsystems involved. These failures together with the reduced availability ensuing from them will be simulated. The example of this study is restricted to reduced availabilities of two subsystems, namely the reactor and the tritium recovery system. For these subsystems the influence of statistically varying intervals of operation is considered. Strategies we selected which are representative of expected modes of operation. In the design of a fuel cycle, care will be taken that prescribed availabilities of the subsystems can be achieved; however, the description of reactor operation is a complex task since operation breaks down into several campaigns for which rules have been specified which enable determination of whether a campaign has been successful and can be stopped. Thus, it is difficult to predict the overall behavior prior to a simulation which includes stochastic elements. Using the example mentioned above the capabilities of the tool will be illustrated; besides the presentation of results of inventory simulation, the applicability of these data will be discussed. (orig.)

  16. The European contribution to the ITER Remote Maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Damiani, C., E-mail: carlo.damiani@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Annino, C.; Balagué, S.; Bates, P.; Ceccanti, F.; Di Mascio, T.; Dubus, G.; Esqué, S.; Gonzalez, C.; Lewczanin, M.; Locke, D.; Mont, L.; Olajos, K.; Ranz, R.; Shuff, R.; Puiu, A.; Van Hille, C.; Van Uffelen, M. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Choi, C.H.; Friconneau, J.P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2014-10-15

    Highlights: •The article introduces the needs for remote maintenance in ITER. •It also discusses some of the issues related to the cultural transition from tokamaks as plasma physics to nuclear reactors. •It highlights the related cultural change and the implications on plant topology and maintenance. •Then, it presents those remote handling systems that will be procured by Europe. •The article emphasises the need of a major involvement of industries from now on. -- Abstract: For a first-of-a-kind nuclear fusion reactor like ITER, remote maintainability of neutron-activated components is one of the key aspects of plant design and operations, and a fundamental ingredient for the demonstration of long-term viability of fusion as energy source. The European Domestic Agency (EU DA, i.e. Fusion for Energy, F4E) is providing important support to the ITER Organisation (IO) in specifying the functional requirements of the Remote Handling (RH) Procurement Packages (i.e. the subsystems allocated to EU DA belonging to the overall ITER Remote Maintenance Systems IRMS), and in performing design and R and D activities – with the support of national laboratories and industries – in order to define a sound concept for these packages. Furthermore, domestic industries are being involved in the subsequent detailed design, validation, manufacturing and installation activities, in order to actually fulfil our procurement-in-kind obligations. After an introduction to ITER Remote Maintenance, this paper will present status and next stages for the RH systems allocated to EU DA, and will also illustrate complementary aspects related to cross cutting technologies like radiation tolerant components and RH control systems. Finally, the way all these efforts are coordinated will be presented together with the overall implementation scenario and key milestones.

  17. Technical issues of RAFMs for the fabrication of ITER Test Blanket Module

    International Nuclear Information System (INIS)

    Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki

    2007-01-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems, as it has they have been developed based on massive industrial experience of ferritic/martensitic steel replacing Mo and Nb of high chromium heat resistant martensitic steels (such as modified 9Cr-1Mo) with W and Ta, respectively. F82H and JLF-1 are RAFMs, which have been developed and studied in Japan and the various effects of irradiation were reported. F82H is designed with emphasis on high temperature property and weldability, and was provided and evaluated in various countries as a part of the IEA fusion materials development collaboration. The JAEA/US collaboration program also has been conducted with the emphasis on irradiation effects of F82H. Now, among the existing database for RAFMs the most extensive one is that for F82H. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of ITER Test Blanket Module (TBM) suggested from the recent achievements in Japan. It is desirable to make the status of RAFMs equivalent to commercial steels to use RAFMs as the ITER-TBM structural material. This would require demonstrating the reproducibility and weldability as well as providing the database. The excellent reproducibility of F82H has been demonstrated with four 5-ton-heats, and two of them were provided as F82H-IEA heats. It has been also proved that F82H could be provided as plates (thickness of 1.5 to 55 mm), pipes and rectangular tubes. It is also important to have the excellent weldability as the TBM has about 300m length of weld line, and it was proved through TIG, EB and YAG weld test performed in air atmosphere. Various mechanical and microstructural data have been accumulated including long-term tests such as creep rupture tests and aging tests. Although F82H is a well-perceived RAFM as the ITER-TBM structural material, some issues are

  18. 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

  19. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m{sup 2} for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m{sup 2} for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  20. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    International Nuclear Information System (INIS)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m 2 for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m 2 for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  1. Japanese contributions to containment structure, assembly and maintenance and reactor building for ITER

    International Nuclear Information System (INIS)

    Shibanuma, Kiyoshi; Honda, Tsutomu; Kanamori, Naokazu

    1991-06-01

    Joint design work on Conceptual Design Activity of International Thermonuclear Experimental Reactor (ITER) with four parties, Japan, the United States, the Soviet Union and the European Community began in April 1988 and was successfully completed in December 1990. In Japan, the home team was established in wide range of collaboration between JAERI and national institute, universities and heavy industries. The Fusion Experimental Reactor (FER) Team at JAERI is assigned as a core of the Japanese home team to support the joint Team activity and mainly conducted the design and R and D in the area of containment structure, remote handling and plant system. This report mainly describes the Japanese contribution on the ITER containment structure, remote handling and reactor building design. Main areas of contributions are vacuum vessel, attaching locks, electromagnetic analysis, cryostat, port and service line layout for containment structure, in-vessel handling equipment design and analysis, blanket handling equipment design and related short term R and D for assembly and maintenance, and finally reactor building design and analysis based on the equipment and service line layout and components flow during assembly and maintenance. (author)

  2. Progress of R&D on water cooled ceramic breeder for ITER test blanket system and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Yoshinori, E-mail: kawamura.yoshinori@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Sato, Satoshi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Konno, Chikara; Edao, Yuki; Hayashi, Takumi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Yamanishi, Toshihiko [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • Thermo-hydraulic calculation in the TBM at the water ingress event has been done. • Shielding calculations for the ITER equatorial port #18 were conducted by using C-lite model. • Prototypic pebbles of Be{sub 17}Ti{sub 2} and Be{sub 12}V had a good oxidation property similar to Be{sub 12}Ti pebble. • Li rich Li{sub 2}TiO{sub 3} pebbles were successfully fabricated using the emulsion method by controlling sintering atmosphere. • New tritium production/recovery experiments at FNS have been started by using ionization chamber as on-line gas monitor. - Abstract: The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

  3. A robust helium-cooled shield/blanket design for ITER

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Bourque, R.F.; Baxi, C.B.

    1993-11-01

    General Atomics Fusion and Reactor Groups have completed a helium-cooled, conceptual shield/blanket design for ITER. The configuration selected is a pressurized tubes design embedded in radially oriented plates. This plate can be made from ferritic steel or from V-alloy. Helium leakage to the plasma chamber is eliminated by conservative, redundant design and proper quality control and inspection programs. High helium pressure at 18 MPa is used to reduce pressure drop and enhance heat transfer. This high gas pressure is believed practical when confined in small diameter tubes. Ample industrial experience exists for safe high gas pressure operations. Inboard shield design is highlighted in this study since the allowable void fraction is more limited. Lithium is used as the thermal contacting medium and for tritium breeding, its safety concerns are minimized by a modular, low inventory design that requires no circulation of the liquid metal for the purpose of heat removal. This design is robust, conservative, reliable, and meets all design goals and requirements. It can also be built with present-day technology

  4. Laser welding to expand the allowable gap in bore welding for ITER blanket hydraulic connection

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, Hisashi, E-mail: tanigawa.hisashi@jaea.go.jp; Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

    2015-10-15

    For application to bore welding of hydraulic connection in the ITER blanket module, laser welding presents the following benefits: low weld heat input is preferred for re-welding of the irradiated material. Its contactless process can intrinsically avoid a failure mode of the tool sticking on the weld. The exact requirements for pipe alignment were assessed in comparison with the assembly tolerance. The groove geometry was modified to expand the allowable initial gap. The groove was machined to be partially thick to obviate the filler wire. First, plates with partially thick grooves were welded to elucidate the preferred groove geometry and welding conditions. With the modified groove, the plates were welded for the initial gap of 1.0 mm. Then the groove geometry and welding conditions were adjusted based on results of pipe welding tests. By application of the additional 0.5-mm-thick and 2.5-mm-wide metal in the groove, pipes with an initial gap of 0.7 mm were welded successfully.

  5. Progress on the Fabrication Methods Development for the Korean Test Blanket Module First Wall in the ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Cho, Seung Yon

    2010-01-01

    A Korean helium cooled molten lithium (HCML) test blanket module (TBM) has been designed to be tested in the International Thermonuclear Experimental Reactor (ITER) TBM and related fabrication methods have been developed especially for the purpose of joining. Since the first wall (FW) of the HCML TBM is composed of a beryllium (Be) as an armor material and a FMS as a structural one, joining with Be to FMS and FMS to FMS should be developed in order to fabricate it

  6. Assessment of tritiated activities in the radwaste generated from ITER Chinese helium cooled ceramic breeding test blanket module system

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chang An, E-mail: chenchangan@caep.cn; Liu, Lingbo; Wang, Bo; Xiang, Xin; Yao, Yong; Song, Jiangfeng

    2016-11-15

    Highlights: • Approaches were developed for calculation/evaluation of tritium activities in the materials and components of a TBM system, with tritium permeation being considered for the first time. • Almost all tritiated materials and components were considered in CNHCCB TBM system including the TBM set, connection pipes, and the ancillary tritium handling systems. • Tritium activity data in HCCB TBM system were updated. Some of which in directly tritium contacted components are to be 2 or 4 magnitudes higher than the original neutron transmutation calculations. • The radwaste amount from both operation and decommission of HCCB TBM system was evaluated. - Abstract: Chinese Helium Cooled Ceramic Breeding Test blanket Module (CNHCCB TBM) will be tested in the ITER machine for the feasibility of in pile tritium production for a future magnetic confinement fusion reactor. The tritium inventories/retentions in the material/components were evaluated and updated mainly based on the tritium diffusion/permeation theory and the analysis of some reported data. Tritiated activities rank from less than 10 Bq g{sup −1} to 10{sup 9} Bq g{sup −1} for the different materials or components, which are generally higher than those from the previous neutron transmutation calculation. The amounts of tritiated radwaste were also estimated according to the operation, decommission, maintenance and replacement strategies, which vary from several tens of kilograms to tons in the different operation phases. The data can be used both for the tritium radiological safety evaluation and radwaste management of CNHCCB TBM set and its ancillary systems.

  7. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 1

    International Nuclear Information System (INIS)

    Malang, S.; Reimann, J.; Sebening, H.; Barleon, L.; Bogusch, E.; Bojarsky, E.; Borgstedt, H.U.; Buehler, L.; Casal, V.; Deckers, H.; Feuerstein, H.; Fischer, U.; Frees, G.; Graebner, H.; John, H.; Jordan, T.; Kramer, W.; Krieg, R.; Lenhart, L.; Malang, S.; Meyder, R.; Norajitra, P.; Reimann, J.; Schwenk-Ferrero, A.; Schnauder, H.; Stieglitz, R.; Oschinski, J.; Wiegner, E.

    1991-12-01

    A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary, Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated R and D-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required R and D-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.) [de

  8. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 2. Detailed version

    International Nuclear Information System (INIS)

    John, H.; Malang, S.; Sebening, H.

    1991-12-01

    A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary. Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated RandD-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required RandD-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.) [de

  9. Numerical simulation of the transient thermal-hydraulic behaviour of the ITER blanket cooling system under the draining operational procedure

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    Highlights: • ITER blanket cooling system hydraulic behaviour is studied under draining transient. • A computational approach based on the finite volume method has been followed. • Draining efficiency has been assessed in term of transient duration and residual water. • Transient duration ranges from ∼40 to 50 s, under the reference draining scenario. • Residual water is predicted to range from few tens of gram up to few kilograms. - Abstract: Within the framework of the research and development activities supported by the ITER Organization on the blanket system issues, an intense analysis campaign has been performed at the University of Palermo with the aim to investigate the thermal-hydraulic behaviour of the cooling system of a standard 20° sector of ITER blanket during the draining transient operational procedure. The analysis has been carried out following a theoretical-computational approach based on the finite volume method and adopting the RELAP5 system code. In a first phase, attention has been focused on the development and validation of the finite volume models of the cooling circuits of the most demanding modules belonging to the standard blanket sector. In later phase, attention has been put to the numerical simulation of the thermal-hydraulic transient behaviour of each cooling circuit during the draining operational procedure. The draining procedure efficiency has been assessed in terms of both transient duration and residual amount of coolant inside the circuit, observing that the former ranges typically between 40 and 120 s and the latter reaches at most ∼8 kg, in the case of the cooling circuit of twinned modules #6–7. Potential variations to operational parameters and/or to circuit lay-out have been proposed and investigated to optimize the circuit draining performances. In this paper, the set-up of the finite volume models is briefly described and the key results are summarized and critically discussed.

  10. 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)

  11. 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)

  12. Manufacturing and testing of full scale prototype for ITER blanket shield block

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sa-Woong, E-mail: swkim12@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Duck-Hoi; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Sung-Ki [WONIL Co., Ltd., Haman (Korea, Republic of); Kang, Sung-Chan [POSCO Specialty Steel Co., Ltd., Changwon (Korea, Republic of); Zhang, Fu; Kim, Byoung-Yoon [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ahn, Hee-Jae; Lee, Hyeon-Gon; Jung, Ki-Jung [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-04-15

    Highlights: • 316L(N)-IG forged steel was successfully fabricated and qualified. • Related R&D activities were implemented to resolve the fabrication issues. • SB #8 FSP was successfully manufactured with conventional fabrication techniques. • All of the validation tests were carried out and met the acceptance criteria. - Abstract: Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program. 316L(N)-IG stainless steel forging blocks with 1.80L × 1.12W × 0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues. This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D.

  13. Development work for the manufacture of a blanket shield prototype for ITER

    International Nuclear Information System (INIS)

    Boireau, B.; Boudot, C.; Cottin, A.; Lorenzetto, P.; Jacquinot, F.; Bucci, P.; Gillia, O.; Vidotto, F.

    2006-01-01

    In the frame of the blanket development for ITER, an R-and-D programme was implemented for the manufacture of a shield prototype by powder Hot Isostatic Pressing (HIPping). The shield consists of a Stainless steel forged block drilled and machined, at the back of which 3D bent tubes are HIPped inside a powder layer. This paper describes the development work through the manufacturing of several mock ups that leads us to be confident for the shield prototype manufacturing. The paper is divided into 2 parts, the first one related to the machining development and validation, the second one relating to the HIP development and validation. A partial full scale mock up for the machining development (machined PFSMU) was manufactured with no particular problems, all the main identified difficulties in machining like deep drilling and castellation machining where overcome and the mock up was conform to the specification. The manufacturing of a HIPped PFSMU for the HIP development was done after the manufacturing of smaller mock-ups each representing a particular detailed design point. A computer simulation work gave us some design recommendation, and the compared analysis of the numerical simulation and experimental results lead us to predict the distortions on the PFSMU HIPped mock up. The HIPped distortions that were the main uncertainty were assessed through small mock ups and bigger one. The mechanical characteristics of the joints are conform to the specification. Associated to the mock up manufacturing is the ultrasonic test development which consists in designing and manufacturing a miniaturized probe travelling inside the bent tube after the HIP cycle to examine the joint tube / powder among others. This ultrasonic development allowed the examination of the HIPped PFSMU mock up that concluded this development work. (author)

  14. Verification of dimensional stability on ITER blanket shield block after stress relieving

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sa-Woong, E-mail: swkim12@nfri.re.kr; Jung, Hun-Chea; Ha, Min-Su; Shim, Hee-Jin

    2016-11-01

    Highlights: • The SB#08 FSP were manufactured by using conventional manufacturing processes such as cutting, milling, drilling and welding. • Especially, a strong back system was adopted in order to prevent welding deformation during cover plate welding process. • Post-Welding Heat Treatment (PWHT) for stress relieving and Hot He Leak Test (HHLT) were waived from the lake of huge test facility in the pre-qualification program. • The PWHT combined with the HHLT, however, were implemented to remove the residual stress and to confirm the soundness of welded parts as an internal R&D activities after the pre-qualification program. • Three dimensional inspection also carried out after the PWHT to check the dimensional stabilization. - Abstract: The tight tolerance requirement is one of key issue to manufacture the ITER blanket shield blocks (SBs) which have many interfaces with the First Wall (FW) and Vacuum Vessel (VV). Manufactured SB shall be satisfied with general tolerances (Class “C” of ISO 2768-1 and “L” of ISO 2768-2) and specific tolerance in 2D general assembly drawings. In order to fulfill the tight tolerance requirements in the final stage of SB, stress relieving after welding operations in the manufacturing process shall be performed. Hot helium leak test, Post Welding Heat Treatment (PWHT) and three-dimensional inspection before and after heat treatment were implemented by using the Full Scale Prototype (FSP) of SB in the framework of domestic R&D activities. The hot He leak test was performed at 250 °C for 30 min, and the result was satisfied the requirements. PWHT was carried out at 400 °C for 24 h by brazing furnace with test chamber. The deformation value before and after was measured by contact type coordinate measuring machine. The objective of this study is to verify dimensional stability of SB after stress relieving. The results will support to determine the machining allowance prior to welding process.

  15. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

    Strebkov, Yu.; Sidorov, A.; Danilov, I.

    1994-01-01

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

  16. Prototyping of the Blanket Shield Module for the ITER EC H and CD Upper launcher

    Energy Technology Data Exchange (ETDEWEB)

    Spaeh, Peter, E-mail: peter.spaeh@kit.edu [KIT – Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, G. [KIT – Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Binni, A. [MAN Diesel and Turbo SE, Deggendorf (Germany); Gessner, R. [KIT – Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Goldmann, A. [MAN Diesel and Turbo SE, Deggendorf (Germany); Grossetti, G. [KIT – Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Kroiss, A. [MAN Diesel and Turbo SE, Deggendorf (Germany); Meier, A. [KIT – Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Obermeier, C. [MAN Diesel and Turbo SE, Deggendorf (Germany); Scherer, T.; Schreck, S.; Strauss, D.; Vaccaro, A. [KIT – Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2014-10-15

    Highlights: • ITER EC H and CD prototype of structural In-vessel components manufactured and analyzed. • Preliminary design was adapted according to manufacturing requirements. • Analysis of flow characteristics for cooling system has been performed. Design was optimized according to this analysis. - Abstract: The design of the ITER Electron Cyclotron Heating and Current Drive (ECH and CD) Upper launcher is recently in the first of two final design phases. The first phase deals with the finalization of all FCS (First Confinement System) components as well as with specific design progress for the remaining In-vessel components. The most outstanding structural In-vessel component of an ECH and CD Upper launcher is the Blanket Shield Module (BSM) with the First Wall Panel (FWP). Both of them form the plasma facing part of the launcher, which has to meet strong demands on dissipation of nuclear heat loads and mechanical rigidity. Nuclear heat loads from 3 MW/m{sup 3} at the First Wall Panel’ surface, decaying down to a tenth in a distance of 0.5 m behind of it will affect the BSM and the FWP. Additional heating of maximum 0.5 MW/m{sup 2} due to plasma radiation must be dissipated from the FWP. To guarantee save and homogenous removal of such extensive heat loads, the BSM is designed as a welded steel-case with specific cooling channels inside its wall structure. Attached to its face side is the FWP with a high-power cooling structure. Based on computational analysis the optimum cooling channel geometry has been investigated. Specific pre-prototype tests have been made and associated assembly parameters have been determined in order to identify optimum manufacturing processes and joining techniques, which guarantee a robust design with maximum geometrical accuracy. This paper describes the design, manufacturing and testing of a full-size mock-up of the BSM. The study was carried out in an industrial cooperation with MAN Diesel and Turbo SE.

  17. Prototyping of the Blanket Shield Module for the ITER EC H and CD Upper launcher

    International Nuclear Information System (INIS)

    Spaeh, Peter; Aiello, G.; Binni, A.; Gessner, R.; Goldmann, A.; Grossetti, G.; Kroiss, A.; Meier, A.; Obermeier, C.; Scherer, T.; Schreck, S.; Strauss, D.; Vaccaro, A.

    2014-01-01

    Highlights: • ITER EC H and CD prototype of structural In-vessel components manufactured and analyzed. • Preliminary design was adapted according to manufacturing requirements. • Analysis of flow characteristics for cooling system has been performed. Design was optimized according to this analysis. - Abstract: The design of the ITER Electron Cyclotron Heating and Current Drive (ECH and CD) Upper launcher is recently in the first of two final design phases. The first phase deals with the finalization of all FCS (First Confinement System) components as well as with specific design progress for the remaining In-vessel components. The most outstanding structural In-vessel component of an ECH and CD Upper launcher is the Blanket Shield Module (BSM) with the First Wall Panel (FWP). Both of them form the plasma facing part of the launcher, which has to meet strong demands on dissipation of nuclear heat loads and mechanical rigidity. Nuclear heat loads from 3 MW/m 3 at the First Wall Panel’ surface, decaying down to a tenth in a distance of 0.5 m behind of it will affect the BSM and the FWP. Additional heating of maximum 0.5 MW/m 2 due to plasma radiation must be dissipated from the FWP. To guarantee save and homogenous removal of such extensive heat loads, the BSM is designed as a welded steel-case with specific cooling channels inside its wall structure. Attached to its face side is the FWP with a high-power cooling structure. Based on computational analysis the optimum cooling channel geometry has been investigated. Specific pre-prototype tests have been made and associated assembly parameters have been determined in order to identify optimum manufacturing processes and joining techniques, which guarantee a robust design with maximum geometrical accuracy. This paper describes the design, manufacturing and testing of a full-size mock-up of the BSM. The study was carried out in an industrial cooperation with MAN Diesel and Turbo SE

  18. The thermo-mechanical design of the water cooled PB-17Li test blanket module for ITER

    International Nuclear Information System (INIS)

    Nardi, C.; Palmieri, A.; Pinna, T.; Porfini, M.T.; Rapisarda, M.; Roccella, M.; Futterer, M.; Lucca, F.

    1998-01-01

    The Water Cooled Lithium Lead (WCLL) blanket is one of the two European concepts to be further developed. A Test Blanket Module (TBM) representative of the DEMO blanket shall be tested in ITER. This paper reports on the activities related to the thermo-mechanical design analysis, taking into account the electromagnetic and neutronic loads in normal and off normal conditions. These loads were applied to a finite elements model of the structure, and the structural response was compared to the allowable value, dependent on the operating conditions. Besides the loads assumed by the design specifications (pressure, temperature, etc), electro-mechanical and thermal loads have been evaluated. A model of the TBM has been performed to compute the loads related to the electromagnetic effects of a centered plasma disruption. The thermal loads have been evaluated considering the heat deposition from the plasma and from the neutrons. The neutronic analysis has been carried out also in order to evaluate the shielding characteristics of the TBM. Taking into account the thermal and mechanical loads a fracture mechanics analysis has been carried out. From this analysis the J Ic parameter was evaluated at the crack tip and compared with the allowable value. The work carried out showed that the TBM present design fulfills ITER normal operation requirements. (authors)

  19. Studies on Flat Sandwich-type Self-Powered Detectors for Flux Measurements in ITER Test Blanket Modules

    Science.gov (United States)

    Raj, Prasoon; Angelone, Maurizio; Döring, Toralf; Eberhardt, Klaus; Fischer, Ulrich; Klix, Axel; Schwengner, Ronald

    2018-01-01

    Neutron and gamma flux measurements in designated positions in the test blanket modules (TBM) of ITER will be important tasks during ITER's campaigns. As part of the ongoing task on development of nuclear instrumentation for application in European ITER TBMs, experimental investigations on self-powered detectors (SPD) are undertaken. This paper reports the findings of neutron and photon irradiation tests performed with a test SPD in flat sandwich-like geometry. Whereas both neutrons and gammas can be detected with appropriate optimization of geometries, materials and sizes of the components, the present sandwich-like design is more sensitive to gammas than 14 MeV neutrons. Range of SPD current signals achievable under TBM conditions are predicted based on the SPD sensitivities measured in this work.

  20. Cassette blanket and vacuum building: key elements in fusion reactor maintenance

    International Nuclear Information System (INIS)

    Werner, R.W.

    1977-01-01

    The integration of two concepts important to fusion power reactors is discussed. The first concept is the vacuum building which improves upon the current fusion reactor designs. The second concept, the use of the cassette blanket within the vacuum building environment, introduces four major improvements in blanket design: cassette blanket module, zoning concept, rectangular blanket concept, and internal tritium recovery

  1. Lithium-vanadium advanced blanket development. ITER final report on U.S. contribution: Task T219/T220

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Mattas, R.F. [comps.

    1997-07-01

    The objective of this task is to develop the required data base and demonstrate the performance of a liquid lithium-vanadium advanced blanket design. The task has two main activities related to vanadium structural material and liquid lithium system developments. The vanadium alloy development activity included four subtasks: (1.1) baseline mechanical properties of non irradiated base metal and weld metal joints; (1.2) compatibility with liquid lithium; (1.3) material irradiation tests; and (1.4) development of material manufacturing and joining methods. The lithium blanket technology activity included four subtasks: (2.1) electrical insulation development and testing for liquid metal systems; (2.2) MHD pressure drop and heat transfer study for self-cooled liquid metal systems; (2.3) chemistry of liquid lithium; and (2.4) design, fabrication and testing of ITER relevant size blanket mockups. A summary of the progress and results obtained during the period 1995 and 1996 in each of the subtask areas is presented in this report.

  2. Lithium-vanadium advanced blanket development. ITER final report on U.S. contribution: Task T219/T220

    International Nuclear Information System (INIS)

    Smith, D.L.; Mattas, R.F.

    1997-07-01

    The objective of this task is to develop the required data base and demonstrate the performance of a liquid lithium-vanadium advanced blanket design. The task has two main activities related to vanadium structural material and liquid lithium system developments. The vanadium alloy development activity included four subtasks: (1.1) baseline mechanical properties of non irradiated base metal and weld metal joints; (1.2) compatibility with liquid lithium; (1.3) material irradiation tests; and (1.4) development of material manufacturing and joining methods. The lithium blanket technology activity included four subtasks: (2.1) electrical insulation development and testing for liquid metal systems; (2.2) MHD pressure drop and heat transfer study for self-cooled liquid metal systems; (2.3) chemistry of liquid lithium; and (2.4) design, fabrication and testing of ITER relevant size blanket mockups. A summary of the progress and results obtained during the period 1995 and 1996 in each of the subtask areas is presented in this report

  3. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

  4. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

  5. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 2

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

    1991-10-01

    The BOT (Breeder Outside Tube) Helium Cooled Solid Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test blankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of the test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.) [de

  6. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  7. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Manly, W.D. [Oak Ridge National Lab., TN (United States); Dombrowski, D.E. [Brush Wellman, Inc., Cleveland, OH (United States)] [and others

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers.

  8. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Manly, W.D.; Dombrowski, D.E.

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers

  9. The ITER EC H and CD Upper Launcher: Analysis of vertical Remote Handling applied to the BSM maintenance

    International Nuclear Information System (INIS)

    Grossetti, Giovanni; Aiello, Gaetano; Heemskerk, Cock; Elzendoorn, Ben; Geßner, Robby; Koning, Jarich; Meier, Andreas; Ronden, Dennis; Späh, Peter; Scherer, Theo; Schreck, Sabine; Strauß, Dirk; Vaccaro, Alessandro

    2013-01-01

    This paper deals with Remote Handling activities foreseen on the Blanket Shield Module, the plasma facing component of the ITER Electron Cyclotron Heating and Current Drive Upper Launcher. The maintenance configuration considered here is the Vertical Remote Handling, meaning gravity acting along the launcher radial axis. The plant, where the maintenance under consideration is occurring, is the Hot Cell Facility Work Cell. The study here reported has been carried out within the presently ongoing EFDA Goal Oriented Training program on Remote Handling (GOT-RH), which aims to support ITER activities. This document and its contents have to be considered as part of a more vast RAMI analysis to be developed within the GOT-RH, which aims to maximize the Electron Cyclotron Heating and Current Drive system availability. The Baseline CAD model of the Electron Cyclotron Heating and Current Drive Upper Launcher is currently in its preliminary design phase and does not provide enough details for developing a fully detailed maintenance strategy. Therefore, through a System Engineering approach, a set of assumptions was conceived on the launcher structure, as a basis for development of a Remote Handling strategy. Moreover, to compare different design solutions related to the possibility of integrating a quasi-optical component into the Blanket Shield Module, a Trade-Off was made, and its contents are shown here. The outcome of this System Engineering approach has been formalized into Task Definition Forms whose contents are reported here. The Remote Handling strategy presented in this work will be tested in the near future both through Virtual Reality simulations and through prototype experiments

  10. The ITER EC H and CD Upper Launcher: Analysis of vertical Remote Handling applied to the BSM maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Grossetti, Giovanni, E-mail: giovanni.grossetti@kit.edu [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Heemskerk, Cock [Heemskerk Innovative Technology, Merelhof 2, 2172 HZ Sassenheim (Netherlands); Elzendoorn, Ben [FOM Institute DIFFER, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Geßner, Robby [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Koning, Jarich [Heemskerk Innovative Technology, Merelhof 2, 2172 HZ Sassenheim (Netherlands); Meier, Andreas [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Ronden, Dennis [FOM Institute DIFFER, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Späh, Peter; Scherer, Theo; Schreck, Sabine; Strauß, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2013-10-15

    This paper deals with Remote Handling activities foreseen on the Blanket Shield Module, the plasma facing component of the ITER Electron Cyclotron Heating and Current Drive Upper Launcher. The maintenance configuration considered here is the Vertical Remote Handling, meaning gravity acting along the launcher radial axis. The plant, where the maintenance under consideration is occurring, is the Hot Cell Facility Work Cell. The study here reported has been carried out within the presently ongoing EFDA Goal Oriented Training program on Remote Handling (GOT-RH), which aims to support ITER activities. This document and its contents have to be considered as part of a more vast RAMI analysis to be developed within the GOT-RH, which aims to maximize the Electron Cyclotron Heating and Current Drive system availability. The Baseline CAD model of the Electron Cyclotron Heating and Current Drive Upper Launcher is currently in its preliminary design phase and does not provide enough details for developing a fully detailed maintenance strategy. Therefore, through a System Engineering approach, a set of assumptions was conceived on the launcher structure, as a basis for development of a Remote Handling strategy. Moreover, to compare different design solutions related to the possibility of integrating a quasi-optical component into the Blanket Shield Module, a Trade-Off was made, and its contents are shown here. The outcome of this System Engineering approach has been formalized into Task Definition Forms whose contents are reported here. The Remote Handling strategy presented in this work will be tested in the near future both through Virtual Reality simulations and through prototype experiments.

  11. Verifying elementary ITER maintenance actions with the MS2 benchmark product

    NARCIS (Netherlands)

    Heemskerk, C. J. M.; Elzendoorn, B. S. Q.; Magielsen, A. J.; Schropp, G. Y. R.

    2011-01-01

    A new facility has been taken in operation to investigate the influence of visual and haptic feedback on the performance of remotely executed ITER RH maintenance tasks. A reference set of representative ITER remote handling maintenance tasks was included the master slave manipulator system (MS2)

  12. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 1

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

    1991-10-01

    The BOT (Breeder Outside Tube) Helium Cooled Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test plankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.) [de

  13. Experiment and analysis of hypervapotron mock-ups for preparing the 2nd qualification of the ITER blanket first wall

    International Nuclear Information System (INIS)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Bang, In Cheol

    2010-01-01

    According to the increased heat flux condition up to 5 MW/m 2 in the International Thermonuclear Experimental Reactor (ITER), new design of the blanket first wall (FW) has been considered and the analysis was performed with ANSYS-CFX for checking its temperature with the ITER operation conditions. And a semi-prototype of the FW was proposed to be tested with the similar heat flux conditions under the second qualification for the FW procurement. In order to investigate the fabrication procedure and analysis capability of the code, two types of mock-up were fabricated according to the current semi-prototype design except for bending shape; one with hypervapotron and another without it. They were tested with KoHLT-2 (Korea Heat Load Test) facility and the results were compared with the ones by CFX code. The mass flow rate of inlet coolant was the same as the ITER condition and heat flux was loaded up to 0.48 MW/m 2 heat flux. The results show that the temperature of the mock-up can be predicted using the CFX code even with the complex geometry and the hypervapotron shows its function to increase the cooling.

  14. Non-LTE line-blanketed model atmospheres of hot stars. 1: Hybrid complete linearization/accelerated lambda iteration method

    Science.gov (United States)

    Hubeny, I.; Lanz, T.

    1995-01-01

    A new munerical method for computing non-Local Thermodynamic Equilibrium (non-LTE) model stellar atmospheres is presented. The method, called the hybird complete linearization/accelerated lambda iretation (CL/ALI) method, combines advantages of both its constituents. Its rate of convergence is virtually as high as for the standard CL method, while the computer time per iteration is almost as low as for the standard ALI method. The method is formulated as the standard complete lineariation, the only difference being that the radiation intensity at selected frequency points is not explicity linearized; instead, it is treated by means of the ALI approach. The scheme offers a wide spectrum of options, ranging from the full CL to the full ALI method. We deonstrate that the method works optimally if the majority of frequency points are treated in the ALI mode, while the radiation intensity at a few (typically two to 30) frequency points is explicity linearized. We show how this method can be applied to calculate metal line-blanketed non-LTE model atmospheres, by using the idea of 'superlevels' and 'superlines' introduced originally by Anderson (1989). We calculate several illustrative models taking into accont several tens of thosands of lines of Fe III to Fe IV and show that the hybrid CL/ALI method provides a robust method for calculating non-LTE line-blanketed model atmospheres for a wide range of stellar parameters. The results for individual stellar types will be presented in subsequent papers in this series.

  15. Analysis of the HCPB breeder blanket bock-up experiment for ITER using SUSD3D code

    International Nuclear Information System (INIS)

    Kodeli, I.

    2005-01-01

    In order to validate new nuclear cross-section evaluations, method development and design of the helium-cooled pebble bed (HCPB) test blanket module of ITER a benchmark experiment was performed this year at the Frascati Neutron Generator (FNG) in the scope of the EFF (European Fusion File) project in Europe. The objective of this experiment is to study the tritium breeding ratio and other nuclear quantities in a breeder blanket in order to establish and improve the quality of related JEFF nuclear data. The experiment consists of a metallic beryllium set-up with two double layers of breeder material (Li 2 CO 3 powder). The reaction rate measurements include the Li 2 CO 3 pellets (tritium breeding ratio), activation foils, and neutron and gamma spectrometers inserted at several axial and lateral locations in the block. Our task is to perform the deterministic transport, and cross section sensitivity and uncertainty analysis. The role of the cross-section sensitivity and uncertainty analysis is to optimise the design of the benchmark, and to assist in the interpretation of the measurement results. The paper presents the pre- and post- analysis of the benchmark experiment. (author)

  16. Investigation of effect of post weld heat treatment conditions on residual stress for ITER blanket shield blocks

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hun-Chea, E-mail: hcjung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Sa-Woong [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Yun-Hee [Division of Convergence Technology, Korea Research Institute of Standard and Science (KRISS), Daejeon (Korea, Republic of); Baek, Seung-Wook [Division of Industrial Metrology, Korea Research Institute of Standard and Science (KRISS), Daejeon (Korea, Republic of); Ha, Min-Su; Shim, Hee-Jin [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • PWHT for ITER blanket shield block should be performed for dimensional stability. • Investigation of the effect of PWHT conditions on properties was performed. • Instrumented indentation method for evaluation of properties was used. • Residual stress and hardness decreased with increasing PWHT temperature. • Optimization of PWHT conditions would be needed for satisfaction of requirement. - Abstract: The blanket shield block (SB) shall be required the tight tolerance because SB interfaces with many components, such as flexible support keypads, First Wall (FW) support contact surfaces, FW central bolt, electrical strap contact surfaces and attachment inserts for both FW and Vacuum Vessel (VV). In order to fulfil the tight tolerance requirement, stress relieving shall be performed for dimensional stability after cover welding operation. In this paper, effect of Post Weld Heat Treatment (PWHT) conditions, temperature and holding time, was investigated on the residual stress and hardness. The 316L Stainless Steel (SS) was prepared and welded by manual TIG welding by using filler material with 2.4 mm of diameter. Welded 316L SS plate was machined to prepare the specimen for PWHT. PWHT was implemented at 250, 300, 400 °C for 2 and 3 h (400 °C only) and residual stress after relaxation were determined. The evaluation of residual stress and hardness for each specimen was carried out by instrumented indentation technique. The residual stress and hardness were decreased with increasing the heat treatment temperature and holding time.

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

    NARCIS (Netherlands)

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

    2013-01-01

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

  18. The TBM-CA configuration management approach for the ITER test blanket module - application to the HCLL TBS

    International Nuclear Information System (INIS)

    Jourd'Heuil, L.; Panayotov, D.; Salavy, J.-F.; Storto, C.; Colombo, M.; Sardain, P.

    2011-01-01

    The European Test Blanket Modules (EU-TBM) are first prototypes of a fusion reactor breeding blanket. They will be tested in dedicated equatorial ports n o 16 of ITER. Technical developments are performed by a Consortium of European Associates (TBM-CA) and supported within the framework of F4E agency. Designing a complex nuclear system like TBM for ITER necessitates an organizational structure inside the consortium to manage in permanence the coherence between requirements (F4E technical and management specifications) and the TBM development through their life time. At the present stage, evolutionary nature of the design from the different teams is important. Highest priority is assigned to the Management support and Design Integration Team (MDIT) to perform an efficient control of the Configuration Management (CM). The TBM-CA CM comprises 4 main processes: a) identifying configuration of a product characteristics, including its interfaces (Configuration identification), b) controlling the evolution from agreed baseline (Configuration Control), c) creating the knowledge database in order to manage the information all along the lifecycle of the items (Configuration status accounting) and d) verifying the current configuration status of the items (Audits). CM is then a powerful tool to link the requirements for engineering, safety, quality assurance and test and acceptance activities. The application of the CM approach is illustrated through the case of TBM-HCLL (Helium Cooled Lithium Lead). The result shows that the proposed methodology and tools are suitable and provide quality solution for the items with a complex configuration such as TBM HCLL.

  19. Research and development of remote maintenance equipment for ITER divertor maintenance

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka

    2005-02-01

    To facilitate easy remote maintainability, the ITER divertor is divided into 60 cassettes, which are transported in the toroidal and radial directions for replacement through maintenance ports located every 90 degrees using the divertor remote maintenance equipment such as in- and ex-vessel transporters. The cassette of 25 tons has to be transported and installed in the vacuum vessel with a positioning accuracy less than 2 mm in the limited space of the vacuum vessel and maintenance port under the intense gamma radiation field. Based on these requirements, the following design and tests were performed. (1) Link mechanism was studied to apply to the transportation of the heavy cassette in the restricted space. A compact mechanism with links for transportation of heavy cassette is designed through the optimization of the link angle taking account of space requirement and force efficiency. As a test result, the lifting capacity of 30 tons (larger than the cassette weight of 25 tons) using two link mechanisms has been demonstrated in the limited space. (2) Compact link mechanism was also studied to apply for locking of the cassette through the optimization of the link angle taking account of space requirement and force efficiency. As a test result, the final positioning accuracy of 0.03 mm for the 25 tons-cassette installation on the vacuum vessel from the initial positioning error of 5 mm has been demonstrated, so that the test result satisfies the requirement less than 2 mm using the link mechanisms in the limited space. (3) Sensor-based control using simple sensors such as optical fiber for divertor maintenance was tested using the full-scale mock-up divertor cassette and remote maintenance equipment. As a result, it is found that the positioning accuracy of 0.16 mm has been achieved by the optical fiber sensor and this value is sufficient for sensor-based control. In addition, the maintenance operation has been carried out through the human-machine interface

  20. High heat flux thermal-hydraulic analysis of ITER divertor and blanket systems

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Ioki, K.; Tivey, R.; Krassovski, D.; Kubik, D.

    1998-01-01

    Three separate cooling systems are used for the divertor and blanket components, based mainly on flow routing access and on grouping together components with the highest heat load levels and uncertainties: divertor, limiter/outboard baffle, and primary first wall/inboard baffle. The coolant parameters for these systems are set to accommodate peak heat load conditions with a reasonable critical heat flux (CHF) margin. Material temperature constraints and heat transport system space and cost requirements are also taken into consideration. This paper summarises the three cooling system designs and highlights the high heat flux thermal-hydraulic analysis carried out in converging on the design values for the coolant operating parameters. Application of results from on-going high heat flux R and D and a brief description of future R and D effort to address remaining issues are also included. (orig.)

  1. Current status of technology development for fabrication of Indian Test Blanket Module (TBM) of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, T., E-mail: tjk@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam 603102 (India); Rajendra Kumar, E. [TBM Division, Institute for Plasma Research (IPR), Bhat, Gandhinagar 382428 (India)

    2014-10-15

    Highlights: • Status of technology developments for Indian TBM to be installed in ITER is presented. • Procedure development for EB, laser and laser-hybrid welding of RAFM steel presented. • Filler wires for RAFM steel for TIG, NG-TIG and laser-hybrid welding have been developed. • Feasibility of production of channel plate by HIP technology has been demonstrated. - Abstract: Ever since India decided to install its Lead-Lithium Ceramic Breeder (LLCB) TBM in ITER, various technologies for fabrication of Indian TBM are being pursued by IPR and IGCAR, in collaboration with various research laboratories in India. Welding consumables for joining India specific RAFM steels (IN-RAFMS), procedures for hot isostatic pressing, electron beam welding, laser and laser-hybrid welding have been developed. Considering the complex nature and limited access available for inspection, innovative inspection procedures that involved use of phased array ultrasonic and C-scan imaging are also being pursued. This paper presents the current status of these developments and provides a roadmap for the future activities planned in realizing Indian TBM for testing in ITER.

  2. Studies of the ultrasonic testing scheme on bonding quality in shield blanket of ITER

    International Nuclear Information System (INIS)

    Shi Sichao; Shen Jingling; He Fengqi; Jin Wanping

    2007-01-01

    International Thermonuclear Experimental Reactor (ITER) is an international cooperative item. One of its components, the First Wall (FW) functioning as neutron shielding and cooling, is an important part. According to the component materials, structural features, testing requirements of the FW, and the ultrasonic propagation characteristics, it is suggested that Broad-band ultrasonic can be used to test the bonding quality of the FW. According to the case mentioned above, the Broad-band Ultrasonic Testing scheme was presented, and the ultrasonic testing feasibility was analyzed theoretically in this paper. (authors)

  3. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hee-Jin [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Ha, Min-Su, E-mail: msha12@nfri.re.kr [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Sa-Woong; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Duck-Hoi [ITER Organization, Route de Vinon sur Verdon - CS 90046, 13067 Sant Paul Lez Durance (France)

    2016-11-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K{sub e} factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  4. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    International Nuclear Information System (INIS)

    Shim, Hee-Jin; Ha, Min-Su; Kim, Sa-Woong; Jung, Hun-Chea; Kim, Duck-Hoi

    2016-01-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K_e factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  5. Introducing artificial depth cues to improve task performance in ITER maintenance actions

    NARCIS (Netherlands)

    Heemskerk, C.J.M.; Eendebak, P.T.; Schropp, G.Y.R.; Hermes, H.V.; Elzendoorn, B.S.Q.; Magielsen, A.J.

    2013-01-01

    Maintenance operations on ITER tokamak components will be largely performed by remote handling. In previous work it was shown that representative maintenance tasks could be performed significantly faster with direct visual feedback than with camera feedback. In post-test interviews, operators

  6. 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

  7. Analysis of the ITER ECH Upper Port Launcher remote maintenance using virtual reality

    NARCIS (Netherlands)

    Elzendoorn, B.; Baar, M. de; Chavan, R.; Goodman, T.; Heemskerk, C.; Heidinger, R.; Kleefeldt, K.; Koning, J.; Sanders, S.; Späh, P.; Strauss, D.; Verhoeven, T.; Vreede, F. de

    2009-01-01

    All ITER sub-systems of remote handling (RH) classes 1 and 2 have to be remotely maintainable. The maintenance strategy for these components has to ensure system availability after failure or scheduled maintenance. This paper shows how virtual reality (VR) simulation [1] can be used as a tool to

  8. FMECA about pre-treatment system for purge gas of test blanket module in ITER

    International Nuclear Information System (INIS)

    Fu Wanfa; Luo Deli; Tang Tao

    2012-01-01

    The pre-treatment system for purge gas of TBM will be installed in Port Cell for installing TBM in ITER, the function of which includes filtering purge gas, removing HTO, cooling, and adjusting flow rate, etc. The purge gas treated will be conveyed into TES (Tritium Extraction System). The technological process and system components in pre-treatment system were introduced. Tritium releasing risk was regarded as failure criterion; failure mode, effects and criticality analysis (FMECA) were carried out and several weaknesses or failure mode in the system were found. Besides, risk priority number (RPN) and failure mode criticality were calculated. Finally, some design improvement measures and usage compensation measures were given. At last, four important potential failure modes were found out. The analysis will provide the design basis for reducing risk of excessive tritium releasing, which is also a useful assist for safety analysis about other tritium system. (authors)

  9. Revisiting the analysis of passive plasma shutdown during an ex-vessel loss of coolant accident in ITER blanket

    International Nuclear Information System (INIS)

    Rivas, J.C.; Dies, J.; Fajarnés, X.

    2015-01-01

    Highlights: • We have repeated the safety analysis for the hypothesis of passive plasma shutdown for beryllium evaporation during an ex-vessel LOCA of ITER first wall, with AINA code. • We have performed a sensitivity analysis over some key parameters that represents uncertainties in physics and engineering, to identify cliff edge effects. • The obtained results for the 500 MW inductive scenario, with an ex-vessel LOCA affecting a third of first wall surface are similar to those of previous studies and point to the possibility of a passive plasma shutdown during this safety case, before a serious damage is inflicted to the ITER wall. • The sensitivity analysis revealed a new scenario potentially damaging for the first wall if we increase fusion power and time delay for impurity transport, and decrease fraction of affected first wall area and initial beryllium fraction in plasma. • After studying the 700 MW inductive scenario, with an ex-vessel LOCA affecting 10% of first wall surface, with 0.5% of Be in plasma and a time delay twice the energy confinement time, it was found that affected area of first wall would melt before a passive plasma shutdown occurs. - Abstract: In this contribution, the analysis of passive safety during an ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of ITER has been studied with AINA safety code. In the past, this case has been studied using robust safety arguments, based on simple 0D models for plasma balance equations and 1D models for wall heat transfer. The conclusion was that, after first wall heating up due to the loss of all coolant, the beryllium evaporation in the wall surface would induce a growing impurity flux into core plasma that finally would end in a passive shut down of the discharge. The analysis of plasma-wall transients in this work is based in results from AINA code simulations. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering

  10. Design evaluation of the semi-prototype for the ITER blanket first wall qualification

    International Nuclear Information System (INIS)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Kim, Sun Ho; Hong, Bong Guen; Bang, In Cheol

    2010-01-01

    For the second qualification of the First Wall (FW) procurement of the International Thermonuclear Experimental Reactor (ITER), a semi-prototype of the FW has been designed with increased local surface heat flux up to 5 MW/m 2 . With the given conditions, the new semi-prototype design was simulated with the commercial CFD code, the ANSYS-11. The results show that the semi-prototype temperature exceeds the melting temperature of Be, and the current design is required to be modified. In order to enhance cooling, a hypervapotron was added in the design and an analysis with the same code was performed. The results show that the temperature with the hypervapotron reduced by around 100 o C but it was still higher than the melting temperature of Be. The hypervapotron mock-up was fabricated and tested with a variance of inlet coolant flow rates and heat fluxes of up to 1.75 MW/m 2 using the second Korea Heat Load Test (KoHLT-2) facility, in which heat was loaded by a graphite heater through radiation heating. Wall and coolant temperatures were measured and compared with the simulation results. So far, there is a large difference between the experiments and the simulation, and a next experiment is being prepared.

  11. ITER L 6 equatorial maintenance duct remote handling study

    International Nuclear Information System (INIS)

    Millard, J.

    1996-09-01

    The status and conclusions of a preliminary study of equatorial maintenance duct remote handling is reported. Due to issues with the original duct design a significant portion of the study had to be refocused on equatorial duct layout studies. The study gives an overview of some of the options for design of these ducts and the impact of the design on the equipment to work in the duct. To develop a remote handling concept for creating access through the ducts the following design tasks should be performed: define the operations sequences for equatorial maintenance duct opening and closing; review the remote handling requirements for equatorial maintenance duct opening and closing; design concept for door and pipe handling equipment and to propose preliminary procedures for material handling outsides the duct. 35 figs

  12. Proposal for the award of a blanket contract for the supply, installation and maintenance of the LHC access control system

    CERN Document Server

    2004-01-01

    This document concerns the award of a blanket contract for the supply, installation and maintenance of the LHC access control system. Following a market survey carried out among 134 firms in fifteen Member States, a call for tenders (IT-3026/TS/LHC) was sent on 22 January 2004 to eight firms and eight consortia in six Member States. By the closing date, CERN had received nine tenders from two firms and seven consortia in five Member States. The Finance Committee is invited to agree to the negotiation of a blanket contract with the consortium CEGELEC CENTRE EST (FR) - CEGELEC (NL), the lowest technically compliant bidder, for the supply, installation and maintenance of the LHC access control system for a total amount not exceeding 4 600 000 euros (7 141 000 Swiss francs), subject to revision for inflation from 1 January 2007. The rate of exchange used is that stipulated in the tender. The firm has indicated the following distribution by country of the contract value covered by this adjudication proposal: FR - ...

  13. Development of filler wires for welding of reduced activation ferritic martenstic steel for India's test blanket module of ITER

    International Nuclear Information System (INIS)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K.

    2011-01-01

    Highlights: → Weld microstructure produced by RAFMS filler wires are free from delta ferrite. → Cooling rates of by weld thermal cycles influences the presence of delta ferrite. → Weld parameters modified with higher pre heat temperature and high heat input. → PWHT optimized based on correlation of hardness between base and weld metals. → Optimised mechanical properties achieved by proper tempering of the martensite. - Abstract: Indigenous development of reduced activation ferritic martensitic steel (RAFMS) has become mandatory to India to participate in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFMS is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFMS filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFMS. Purpose of this study is to develop filler wires that can be directly used for both tungsten inert gas welding (TIG) and narrow gap tungsten inert gas welding (NG-TIG), which reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, autogenous welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using TIG process at various heat inputs with a preheat temperature of 250 deg. C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimised to qualify the filler wires without the presence of delta-ferrite in

  14. Development of filler wires for welding of reduced activation ferritic martensitic steel for India's test blanket module of ITER

    International Nuclear Information System (INIS)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K.

    2010-01-01

    Indigenous development of reduced activation ferritic-martensitic (RAFM) steel has become necessary for India as a participant in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFM steel is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFM steel filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFM steel. The purpose of this study is to develop filler wires that can be directly used for both gas tungsten arc welding (GTAW) and for narrow-gap gas tungsten arc welding (NG-GTAW) that reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser-MIG welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using GTAW process at various heat inputs with a preheat temperature of 250 C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some amount of delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimized to qualify the filler wires without the presence of delta-ferrite in the weld metal and with optimized mechanical properties. Results showed that the weld metals are free from delta-ferrite. Tensile properties at ambient temperature and at 500 C are well above the specified values, and are much higher than the base metal values. Ductile Brittle Transition Temperature (DBTT) has been evaluated as -81 C based on the 68 J criteria. The present study highlights the basis and methodology

  15. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)]. E-mail: wongc@fusion.gat.com; Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Sawan, M. [University of Wisconsin, Madison, WI (United States); Dagher, M. [University of California, Los Angeles, CA (United States); Smolentsev, S. [University of California, Los Angeles, CA (United States); Merrill, B. [INEEL, Idaho Falls, ID (United States); Youssef, M. [University of California, Los Angeles, CA (United States); Reyes, S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Sze, D.K. [University of California, San Diego, CA (United States); Morley, N.B. [University of California, Los Angeles, CA (United States); Sharafat, S. [University of California, Los Angeles, CA (United States); Calderoni, P. [University of California, Los Angeles, CA (United States); Sviatoslavsky, G. [University of Wisconsin, Madison, WI (United States); Kurtz, R. [Pacific Northwest Laboratory, Richland, WA (United States); Fogarty, P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Zinkle, S. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Abdou, M. [University of California, Los Angeles, CA (United States)

    2006-02-15

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC{sub f}/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 deg. C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R and D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

  16. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

    2006-07-05

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperture of 700C. We have identified critical issues for the concept, some of which inlude the first wall design, the assessment of MHD effectrs with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time, we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gessner, Robby, E-mail: robby.gessner@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano; Grossetti, Giovanni; Meier, Andreas [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Ronden, Dennis [DIFFER – Dutch Institute for Fundamental Energy Physics, P.O. Box 1207, NL-3430 BE Nieuwegein (Netherlands); Spaeh, Peter; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2013-10-15

    Highlights: ► The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint. ► The bolts were designed as “captive” in order to avoid their accidental removal from the joint. ► The bolted flange connection using two sets of 15 captive bolts (M22 × 2) placed along the sides. ► The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. -- Abstract: The final design of the structural system for the ITER EC H and CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of

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

    International Nuclear Information System (INIS)

    Gessner, Robby; Aiello, Gaetano; Grossetti, Giovanni; Meier, Andreas; Ronden, Dennis; Spaeh, Peter; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro

    2013-01-01

    Highlights: ► The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint. ► The bolts were designed as “captive” in order to avoid their accidental removal from the joint. ► The bolted flange connection using two sets of 15 captive bolts (M22 × 2) placed along the sides. ► The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. -- Abstract: The final design of the structural system for the ITER EC H and CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of

  19. Conceptual design of cesium removal device for ITER NBI maintenance

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

    Cesium is required in order to generate a stable negative ion of hydrogen in an ion source of the neutral beam injector (NBI), which is one of the plasma-heating devices for International Thermonuclear Experimental Reactor (ITER). After long time operation of the NBI, the cesium deposits to the insulators supporting the electrode. Due to the deterioration of the insulation resistance, the continuous operation of the NBI will be difficult. In addition, the NBI device is activated by neutrons from D-T plasma, so that periodic removal and cleaning of the cesium on the insulators by remove handling is required. A study of the cesium removal scenario and the device is therefore required considering remote handling. In this report, a cesium removal procedure and conceptual design of the cesium removal device using laser ablation technique are studied, and the feasibility of the laser ablation method is shown. (author)

  20. Analysis of the ITER ECH Upper Port Launcher remote maintenance using virtual reality

    International Nuclear Information System (INIS)

    Elzendoorn, Ben; de Baar, Marco; Chavan, Rene; Goodman, Timothy; Heemskerk, Cock; Heidinger, Roland; Kleefeldt, Klaus; Koning, Jarich; Sanders, Stephen; Spaeh, Peter; Strauss, Dirk; Verhoeven, Toon; Vreede, Fred de

    2009-01-01

    All ITER sub-systems of remote handling (RH) classes 1 and 2 have to be remotely maintainable. The maintenance strategy for these components has to ensure system availability after failure or scheduled maintenance. This paper shows how virtual reality (VR) simulation can be used as a tool to analyze the maintenance process, to predict the mean time to repair and to ensure the RH compatibility of one ITER sub-system, the Upper Port Launcher (UPL) . Special emphasis is put on the development of RH procedures and the identification of tooling requirements. The possibility to simulate RH logistics and repair actions in an early stage of the design process allows for the identification of those maintenance actions that require dedicated tests in the Launcher Handling Test Facility at Karlsruhe. The VR analysis, together with dedicated mock-up tests will demonstrate the RH compatibility of the UPL plug, provide input to the design of the Port Plug maintenance area in the ITER Hot Cell, and support the development of RH maintenance tooling.

  1. Analysis and validation center for ITER RH maintenance scenarios in a virtual environment

    NARCIS (Netherlands)

    Elzendoorn, B. S. Q.; M.R. de Baar,; Hamilton, D.; Heemskerk, C. J. M.; Koning, J. F.; Ronden, D. M. S.

    2012-01-01

    A facility for detailed simulation of maintenance processes in the ITER Hot Cell Facility (HCF) has been taken into operation. The facility mimics the Remote Handling (RH) work-cells as are presently foreseen. Novel virtual reality (VR) technology, extended with a physics engine is used to create a

  2. Designer's guidebook for first wall/blanket/shield assembly, maintenance, and repair

    International Nuclear Information System (INIS)

    1983-01-01

    This is the initial issue of the guidebook. Since a guidebook of this type must incorporate information concerning a wide range of subjects, much additional data will eventually be included. The guidebook will document, in summary and easily referenceable form, data, designs, design concepts, design guidelines and background information useful to the FWBS and to the Maintenance System designer. In providing guidelines for the AMR of the FWBS, the guidebook must, of necessity, include guidelines for all aspects of maintenance associated with the FWBS. These include most maintenance operations within the reactor room necessary to gain access, identify faults, and handle equipment related to FWBS maintenance. In addition, the guidelines include those required to define facility requirements for handling and repair of FWBS and related reactor components external to the reactor room. Particular emphasis is given to remote maintenance design and operations

  3. Designer's guidebook for first wall/blanket/shield assembly, maintenance, and repair

    Energy Technology Data Exchange (ETDEWEB)

    1983-12-30

    This is the initial issue of the guidebook. Since a guidebook of this type must incorporate information concerning a wide range of subjects, much additional data will eventually be included. The guidebook will document, in summary and easily referenceable form, data, designs, design concepts, design guidelines and background information useful to the FWBS and to the Maintenance System designer. In providing guidelines for the AMR of the FWBS, the guidebook must, of necessity, include guidelines for all aspects of maintenance associated with the FWBS. These include most maintenance operations within the reactor room necessary to gain access, identify faults, and handle equipment related to FWBS maintenance. In addition, the guidelines include those required to define facility requirements for handling and repair of FWBS and related reactor components external to the reactor room. Particular emphasis is given to remote maintenance design and operations.

  4. Major achievements of the European shield blanket R and D during the ITER EDA, and their relevance for future next step machines

    Energy Technology Data Exchange (ETDEWEB)

    Daenner, W. E-mail: daenner@ipp.mpg.de; Cardella, A.; Jones, L.; Lorenzetto, P.; Maisonnier, D.; Malavasi, G.; Peacock, A.; Rodgers, E.; Tavassoli, F

    2000-11-01

    In the frame of the international thermonuclear experimental reactors (ITER) collaboration, the European home team (EU HT) has committed significant efforts on the R and D for the Shield Blanket. This paper summarises the main achievements of this programme, which have been obtained over the last 7 years. The depth of R and D extends from generic activities up to the manufacture of prototypes, but has, in accordance with the design progress, reached different stages of maturity for the various components. New ITER options being considered since early 1998 have not made these activities irrelevant. With few exceptions, the results are still applicable for less ambitious next step machines, or transferable to components with similar functions or requirements.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  7. Water-cooled Pb-17Li test blanket module for ITER: impact of the structural material grade on the neutronic responses

    Energy Technology Data Exchange (ETDEWEB)

    Vella, G.; Aiello, G.; Oliveri, E. [Palermo Univ. (Italy). Dipt. di Ingegneria Nucl.; Fuetterer, M.A.; Giancarli, L. [CEA - Saclay, DRN/DMT/SERMA, Gif-sur-Yvette (France); Tavassoli, F. [CEA - Saclay, CEREM, Gif-sur-Yvette (France)

    1998-10-01

    The water-cooled lithium lead (WCLL) DEMO blanket is one of the two EU lines to be further developed with the aim of manufacturing by 2010 a test blanket module for ITER (TBM). In this paper results of a 3D-Monte Carlo neutronic analysis of the TBM design are reported. A fully 3D heterogeneous model of the WCLL-TBM has been inserted into an existing ITER model accounting for a proper D-T neutron source. The structural material assumed for the calculations was martensitic 9% Cr steel code named Z 10 CDV Nb 9-1. Results have been compared with those obtained using MANET. The main nuclear responses of the TBM have been determined, such as detailed power deposition density, material damage through DPA and He and H gas production rate, radial distribution of tritium production rate and total tritium production in the module. The impact of using natural lithium on the TBM system operation has also been evaluated. (orig.) 13 refs.

  8. Achievements in the development of the Water Cooled Solid Breeder Test Blanket Module of Japan to the milestones for installation in ITER

    International Nuclear Information System (INIS)

    Tsuru, Daigo; Tanigawa, Hisashi; Hirose, Takanori; Mohri, Kensuke; Seki, Yohji; Enoeda, Mikio; Ezato, Koichiro; Suzuki, Satoshi; Nishi, Hiroshi; Akiba, Masato

    2009-01-01

    As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, a water cooled solid breeder (WCSB) TBM is being developed. This paper shows the recent achievements towards the milestones of ITER TBMs prior to the installation, which consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, targeting the detailed design final report in 2012, structure designs of the WCSB TBM and the interfacing components (common frame and backside shielding) that are placed in a test port of ITER and the layout of the cooling system are presented. As for the module qualification, a real-scale first wall mock-up fabricated by using the hot isostatic pressing method by structural material of reduced activation martensitic ferritic steel, F82H, and flow and irradiation test of the mock-up are presented. As for safety milestones, the contents of the preliminary safety report in 2008 consisting of source term identification, failure mode and effect analysis (FMEA) and identification of postulated initiating events (PIEs) and safety analyses are presented.

  9. The use of virtual prototyping and simulation in ITER maintenance device development

    International Nuclear Information System (INIS)

    Mattila, J.; Siuko, M.; Saarinen, H.; Maekinen, H.; Verho, S.; Vilenius, M.; Palmer, J.; Irving, M.

    2006-01-01

    The ITER divertor maintenance takes place approximately every second year. The maintenance occurs in very harsh and mechanically complicated environment. Due to the critical nature of the maintenance operations, the maintenance equipment design and the operation cycle will be verified in DTP2 test platform, in Tampere, Finland. TUT/IHA is working on the ITER divertor maintenance devices. Due to the complexity of the operation environment and tasks to be performed, 3d models and kinematic simulation have been valuable tool when developing the devices. Further, IHA has integrated to the models also dynamic properties of the device, so that it can be discussed as a virtual prototype. The virtual prototype can be used to verify the operation of the device, the operation cycle and also as a platform for developing the control software for the device. For device development, the virtual prototype is used to analyze the dynamic behavior, loading and flexibility of the device. The virtual prototype was also connected to real hardware to verify the operation of one joint. Then, the virtual model in computer was run and the output of the joints was given to a hydraulic cylinder representing disturbance load for an other hydraulic cylinder, which was operating under control software and aiming to move smoothly regardless of the disturbance load. By that way we were able to verify that the real system operates close enough with the simulation model. The virtual model is also used to shorten the time to get the DTP2 platform working. The CMM control software is done with virtual models as ready as possible. The CMM virtual model is connected to one-joint control hardware which allows developing the controller software one joint at time. In this paper, also other possibilities to use virtual prototypes in ITER divertor maintenance development are discussed. (author)

  10. 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.

  11. Strategy for solving a coupled problem of the electromagnetic load analysis and design optimization for local conducting structures to support the ITER blanket development

    Energy Technology Data Exchange (ETDEWEB)

    Rozov, Vladimir, E-mail: vladimir.rozov@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Belyakov, V.; Kukhtin, V.; Lamzin, E.; Mazul, I.; Sytchevsky, S. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation)

    2014-11-15

    Highlights: • We present the way of modeling transient electro-magnetic loads on local conductive domains in the large magnetic system. • Simplification is achieved by decomposing of the problem, multi-scale integral-differential modeling and use of integral parameters. • The intrinsic scale of loads on a localized conductor with eddy is quantified through the load susceptibility tensor. • Solution is searched as response of a simple equivalent dynamic simulator, using control theory methods. • The concept is exemplified with multi-scenario assessment of EM eddy loads on ITER blanket modules. - Abstract: The complexity of the electromagnetic (EM) response of the tokamak structures is one of the key and design-driving issues for the ITER. We consider the specifics of the assessment of ponderomotive forces, acting on local components of a large electro-physical device during electromagnetic transients. A strategy and approach is proposed for the operative EM loads modeling and analysis that enables design optimization at early phases of development. The paper describes a method of principal simplification of the mathematical model, based on the analysis and exploiting specific features and peculiarities of the relevant technical problem, determined by the design and operation of the device and system under consideration. The application of the method for predictive EM loads analysis and corresponding numerical calculations are exemplified for the localized ITER blanket components — shield modules. The example demonstrates the efficiency of EM load analysis in complex electromagnetic systems via a set of simplified models with different scope, contents and level of detail.

  12. Strategy for solving a coupled problem of the electromagnetic load analysis and design optimization for local conducting structures to support the ITER blanket development

    International Nuclear Information System (INIS)

    Rozov, Vladimir; Belyakov, V.; Kukhtin, V.; Lamzin, E.; Mazul, I.; Sytchevsky, S.

    2014-01-01

    Highlights: • We present the way of modeling transient electro-magnetic loads on local conductive domains in the large magnetic system. • Simplification is achieved by decomposing of the problem, multi-scale integral-differential modeling and use of integral parameters. • The intrinsic scale of loads on a localized conductor with eddy is quantified through the load susceptibility tensor. • Solution is searched as response of a simple equivalent dynamic simulator, using control theory methods. • The concept is exemplified with multi-scenario assessment of EM eddy loads on ITER blanket modules. - Abstract: The complexity of the electromagnetic (EM) response of the tokamak structures is one of the key and design-driving issues for the ITER. We consider the specifics of the assessment of ponderomotive forces, acting on local components of a large electro-physical device during electromagnetic transients. A strategy and approach is proposed for the operative EM loads modeling and analysis that enables design optimization at early phases of development. The paper describes a method of principal simplification of the mathematical model, based on the analysis and exploiting specific features and peculiarities of the relevant technical problem, determined by the design and operation of the device and system under consideration. The application of the method for predictive EM loads analysis and corresponding numerical calculations are exemplified for the localized ITER blanket components — shield modules. The example demonstrates the efficiency of EM load analysis in complex electromagnetic systems via a set of simplified models with different scope, contents and level of detail

  13. Comparison of the Tritium permeated from ITER Blanket in normal operation and its short range impact of HT over France, Swiss or Spain

    Energy Technology Data Exchange (ETDEWEB)

    Castro, P.; Velarde, M.; Ardao, J.; Perlado, J.; Sedano, L.; Xiberta, J.

    2015-07-01

    In this paper we assumes the hydrogen isotopes permeation from a liquid metal ITER breeder blanket (assuming normal operation and a LM as DCLL or HCLL blanket) as one of the possible sources of a leak and tritium release,mainly but not only. The paper presents a short range low impact of HT gas activity over France, Swiss or Spain from same cases in 2014 and 2015 releases from ITER. The permeation of hydrogen isotopes is an important experimental issue to take into account into the development of a Tritium Breeder Module for ITER [1]. Tritium cannot be confined -without an uncertainty of 5% in the flux permeation- and therefore HT can be detected (e.g. by ionization chamber) as permeates though the structure of RAFM steel towards the coolant [1]. HT from Pb15.7Li and permeated in Eurofer97 can contaminate the other parts of the system and may be delivered though the normal-vent detritiation system (NVDS). Real time forecast of transport of tritium in air from the fusion reactor towards off-site far downwind though extended tritium clouds into the low levels of the atmosphere is calculated for the short range (up to 24 hours) by the coupling of 2 models the European Centre for Medium Range Weather Forecast (ECMWF) [2] model and the FLEXPART lagrangian dispersion model [3] verified with NORMTRI simulation [4] and implemented in many different cases and scenarios [5, 6, 7]. As a function of daily weather conditions the release will affect just France or already can be delivered towards Swiss when cyclonic circulation, or towards the Iberian Peninsula or Balearic Islands (Spain) when high produce anticyclonic circulation of the air over the Mediterranean Sea. (Author)

  14. e-Learning Application for Machine Maintenance Process using Iterative Method in XYZ Company

    Science.gov (United States)

    Nurunisa, Suaidah; Kurniawati, Amelia; Pramuditya Soesanto, Rayinda; Yunan Kurnia Septo Hediyanto, Umar

    2016-02-01

    XYZ Company is a company based on manufacturing part for airplane, one of the machine that is categorized as key facility in the company is Millac 5H6P. As a key facility, the machines should be assured to work well and in peak condition, therefore, maintenance process is needed periodically. From the data gathering, it is known that there are lack of competency from the maintenance staff to maintain different type of machine which is not assigned by the supervisor, this indicate that knowledge which possessed by maintenance staff are uneven. The purpose of this research is to create knowledge-based e-learning application as a realization from externalization process in knowledge transfer process to maintain the machine. The application feature are adjusted for maintenance purpose using e-learning framework for maintenance process, the content of the application support multimedia for learning purpose. QFD is used in this research to understand the needs from user. The application is built using moodle with iterative method for software development cycle and UML Diagram. The result from this research is e-learning application as sharing knowledge media for maintenance staff in the company. From the test, it is known that the application make maintenance staff easy to understand the competencies.

  15. Verifying elementary ITER maintenance actions with the MS2 benchmark product

    International Nuclear Information System (INIS)

    Heemskerk, C.J.M.; Elzendoorn, B.S.Q.; Magielsen, A.J.; Schropp, G.Y.R.

    2011-01-01

    A new facility has been taken in operation to investigate the influence of visual and haptic feedback on the performance of remotely executed ITER RH maintenance tasks. A reference set of representative ITER remote handling maintenance tasks was included the master slave manipulator system (MS2) benchmark product. The benchmark product was used in task performance tests in a representative two-handed dexterous manipulation test bed at NRG. In the setup, the quality of visual feedback was varied by exchanging direct view with indirect view setups in which visual feedback is provided via video cameras. Interaction forces were measured via an integrated force sensor. The impact of feedback quality on the performance of maintenance tasks at the level of handling individual parts was measured and analysed. Remote execution of the maintenance actions took roughly 3-5 times more time than hands-on. Visual feedback was identified as the dominant factor, including aspects like (lack of) operator control over camera placement, pan, tilt and zoom, lack of 3D perception, image quality, and latency. Haptic feedback was found to be important, but only in specific contact transition and constrained motion tasks.

  16. ITER-FEAT vacuum vessel and blanket design features and implications for the R and D programme

    International Nuclear Information System (INIS)

    Ioki, K.; Cardella, A.; Elio, F.; Onozuka, M.; Daenner, W.; Koizumi, K.; Krylov, V.

    2001-01-01

    A tight fitting configuration of the VV to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the TF ripple. The blanket modules are supported directly by the VV. A full-scale VV sector model has provided critical information related to fabrication technology, and the magnitude of welding distortions and achievable tolerances. This R and D validated the fundamental feasibility of the double-wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and the robustness of solid HIP joining was demonstrated in R and D, by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal. (author)

  17. ITER-FEAT vacuum vessel and blanket design features and implications for the R and D programme

    International Nuclear Information System (INIS)

    Ioki, K.; Cardella, A.; Elio, F.; Onozuka, M.; Daenner, W.; Koizumi, K.; Krylov, V.A.

    2001-01-01

    A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R and D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R and D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal. (author)

  18. ITER-FEAT vacuum vessel and blanket design features and implications for the R&D programme

    Science.gov (United States)

    Ioki, K.; Dänner, W.; Koizumi, K.; Krylov, V. A.; Cardella, A.; Elio, F.; Onozuka, M.; ITER Joint Central Team; ITER Home Teams

    2001-03-01

    A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R&D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal.

  19. Mirror reactor blankets

    International Nuclear Information System (INIS)

    Lee, J.D.; Barmore, W.L.; Bender, D.J.; Doggett, J.N.; Galloway, T.R.

    1976-01-01

    The general requirements of a breeding blanket for a mirror reactor are described. The following areas are discussed: (1) facility layout and blanket maintenance, (2) heat transfer and thermal conversion system, (3) materials, (4) tritium containment and removal, and (5) nuclear performance

  20. Optimization of HIP bonding conditions for ITER shielding blanket/first wall made from austenitic stainless steel and dispersion strengthened copper alloy

    International Nuclear Information System (INIS)

    Sato, S.; Hatano, T.; Kuroda, T.; Furuya, K.; Hara, S.; Enoeda, M.; Takatsu, H.

    1998-01-01

    Optimum bonding conditions were studied on the hot isostatic pressing (HIP) bonded joints of type 316L austenitic stainless steel and dispersion strengthened copper alloy (DSCu) for application to the ITER shielding blanket / first wall. HIP bonded joints were fabricated at temperatures in a 980-1050 C range, and a series of mechanical tests and metallurgical observations were conducted on the joints. Also, bondability of two grades of DSCu (Glidcop Al-25 trademark and Al-15 trademark ) with SS316L was examined in terms of mechanical properties of the HIP bonded joints. From those studies it was concluded that the HIP temperature of 1050 C was an optimal condition for obtaining higher ductility, impact values and fatigue strength. Also, SS316L/Al-15 joints showed better results in terms of ductility and impact values compared with SS316L/Al-25 joints. (orig.)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

  3. Proposal for the award of a blanket purchase contract for the design, supply, installation and maintenance of automatic fire-detection, fire-protection and voice-alarm systems for the Super Proton Synchrotron

    CERN Document Server

    2017-01-01

    Proposal for the award of a blanket purchase contract for the design, supply, installation and maintenance of automatic fire-detection, fire-protection and voice-alarm systems for the Super Proton Synchrotron

  4. Extending Virtual Reality simulation of ITER maintenance operations with dynamic effects

    International Nuclear Information System (INIS)

    Heemskerk, C.J.M.; Baar, M.R. de; Boessenkool, H.; Graafland, B.; Haye, M.J.; Koning, J.F.; Vahedi, M.; Visser, M.

    2011-01-01

    Virtual Reality (VR) simulation can be used to study, improve and verify ITER maintenance operations during preparation. VR can also improve the situational awareness of human operators during actual Remote Handling (RH) operations. Until now, VR systems use geometric models of the environment and the objects being handled and kinematic models of the manipulation systems. The addition of dynamic effects into the VR simulation was investigated. Important dynamic effects are forces due to contact transitions and the bending of beams under heavy loads. A novel dynamics simulation module was developed and introduced as an add-on to the VR4Robots VR software. Tests were performed under simplified test conditions and in the context of realistic ITER maintenance tasks on a benchmark product and on the ECRH Upper Port Launcher Plug (UPL). The introduction of dynamic effects into VR simulations was found to add realism and provide new insights in procedure development. The quality of the haptic feedback depends strongly on the haptic device used to 'display' haptic feedback to the operator. Dynamic effect simulation can also form the basis for real-time guidance support to operators during the execution of maintenance tasks (augmented reality).

  5. ITER diagnostics: Maintenance and commissioning in the hot cell test bed

    International Nuclear Information System (INIS)

    Walker, C.I.; Barnsley, R.; Costley, A.E.; Gottfried, R.; Haist, B.; Itami, K.; Kondoh, T.; Loesser, G.D.; Palmer, J.; Sugie, T.; Tesini, A.; Vayakis, G.

    2005-01-01

    In-vessel diagnostic equipment in ITER integrated in six equatorial and 12 upper ports, 16 divertor cassettes and five lower ports is designed to be removed in modules and then repaired, tested and commissioned in the same location at the ITER hot cell. The repair requirements and tests on these components are described along with design features that facilitate repair. The testing establishes the repair strategy, qualifies the refurbishment work and finally checks the mechanical and diagnostic function before the return of the modules. At the hot cell, a dummy port is provided for tests of mechanical and vacuum integrity as well as commissioning of the diagnostic equipment. The scope of the hot cell maintenance and commissioning activities is summarised and an overview of the integration of the diagnostic equipment is given

  6. Development of laser cutting/welding system for remote maintenance of ITER manifold

    Energy Technology Data Exchange (ETDEWEB)

    Yamaoka, Hiroto; Tsuchiya, Kazuyuki; Awano, Toshihiko [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan); Oka, Kiyoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-09-01

    A remote pipe cutting/welding system using a YAG laser was designed and fabricated for the maintenance of the main structural parts of ITER (International Thermonuclear Experimental Reactor), and a mock-up test carried out. The functions of this system are to cut 100A x Sch 40 pipes of SUS316L by internal access, to adjust the core gap between the as-cut pipe and new pipe, and to weld the pipes automatically. The core gap of the pipes could be decreased within the proper welding conditions by the mock-up test, and sound beads were obtained. (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. An overview of the US work to complete the design of Blanket Shield Modules 7,12 and 13 for the ITER project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Coats, R.; Garde, J.

    2007-01-01

    Part of the US contribution to the ITER project is Blanket Shield Modules (BSM) number 7, 12 and 13 comprising about 20% of the ITER First Wall. These modules are all in the upper half of the vacuum vessel. Starting from outlines of each style of module, general design rules contained in the ITER Design Description Document (DDD), design memos, and CATIA interface drawings, the US Team has worked to fill in the details of the internal water cooling passages, slits to control eddy currents, and methods for efficient and reliable manufacturing of the BSM. Our analysis begins with nuclear heating assessment of complex 3D structures containing water, copper, and steel carried out by the University of Wisconsin using a version of Monte Carlo N-Particle Transport Code (MCNP) that connects directly to CATIA to get the geometry. Computational Fluid Dynamics (CFD) analysis of the coolant flow distribution and pressure drop in a shield module provides the basis for thermal transfer from the BSM to the coolant. The size and position of coolant passages are adjusted to optimize the heat transfer and eliminate hot spots. ITER specified major disruption (MD) and downward vertical disruption (VDE) events are used to calculate the currents induced in the BSM. In this modeling it is necessary to include the vacuum vessel and other BSM near the modules of interest. In order to benchmark the OPERA Electromagnetic Code against the one used for the DDD analysis, we calculated eddy currents and forces on all 18 BSM in a simplified model that matched analysis by Japan. The eddy current forces are used to determine the torque and net force on the BSM. These forces are compared to the load capacity of the mounts and adjustments made to eddy current control slits as needed. Dynamic analysis of the eddy current induced stresses on the BSM and mounts are performed using the ABAQUS code. Static thermal and pressure stresses are calculated using the temperature distributions from CFD analysis

  9. Novel blanket design for ICTR's

    International Nuclear Information System (INIS)

    Abdel-Khalik, S.I.; Conn, R.W.; Wolfer, W.G.; Larsen, E.N.; Sviatoslavsky, I.N.

    1978-01-01

    A novel blanket design for ICTRs is described. This blanket is used in SOLASE, the conceptual laser fusion reactor of the University of Wisconsin. The blanket to be described offers numerous advantages, including low cost, low weight, low induced radioactivity levels, the potential for hands-on maintenance, modular construction, low pressure, ability to decouple first wall and blanket coolant temperatures, adequate breeding, low tritium inventory and leakage, and sufficiently long life

  10. Analysis and validation center for ITER RH maintenance scenarios in a virtual environment

    International Nuclear Information System (INIS)

    Elzendoorn, B.S.Q.; Baar, M. de; Hamilton, D.; Heemskerk, C.J.M.; Koning, J.F.; Ronden, D.M.S.

    2012-01-01

    A facility for detailed simulation of maintenance processes in the ITER Hot Cell Facility (HCF) has been taken into operation. The facility mimics the Remote Handling (RH) work-cells as are presently foreseen. Novel virtual reality (VR) technology, extended with a physics engine is used to create a realistic setting in which a team of Remote Handling (RH) operators can interact with a virtual Hot Cell environment. The physics engine is used to emulate the Hot Cell behavior and to provide tactile feed-back of the (virtual) slave. Multi-operator maintenance scenarios can be developed and tested in virtual reality. Complex interactions between the RH operators and the HCF control system software will be tested. Task performance will be quantified and operational resource consumption will be estimated.

  11. Analysis and validation center for ITER RH maintenance scenarios in a virtual environment

    Energy Technology Data Exchange (ETDEWEB)

    Elzendoorn, B.S.Q., E-mail: B.S.Q.Elzendoorn@rijnhuizen.nl [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE, Nieuwegein (Netherlands); Baar, M. de [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE, Nieuwegein (Netherlands); Hamilton, D. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul-lez-Durance Cedex (France); Heemskerk, C.J.M. [Heemskerk Innovative Technology, Sassenheim (Netherlands); Koning, J.F.; Ronden, D.M.S. [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE, Nieuwegein (Netherlands)

    2012-03-15

    A facility for detailed simulation of maintenance processes in the ITER Hot Cell Facility (HCF) has been taken into operation. The facility mimics the Remote Handling (RH) work-cells as are presently foreseen. Novel virtual reality (VR) technology, extended with a physics engine is used to create a realistic setting in which a team of Remote Handling (RH) operators can interact with a virtual Hot Cell environment. The physics engine is used to emulate the Hot Cell behavior and to provide tactile feed-back of the (virtual) slave. Multi-operator maintenance scenarios can be developed and tested in virtual reality. Complex interactions between the RH operators and the HCF control system software will be tested. Task performance will be quantified and operational resource consumption will be estimated.

  12. Current status of research and development on remote maintenance for fusion components

    International Nuclear Information System (INIS)

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

    2008-01-01

    There is a growing attention to remote maintenance of nuclear fusion reactors. Remote maintenance is planned in ITER tokamak to keep the health of in-vessel components like blankets and divertors. In this article, current status of the development in the remote maintenance equipments and methods, especially for ITER tokamak are reviewed. The newly developed vehicle type and boom type maintenance devices, manipulator, and transfer cask are illustrated. (J.P.N.)

  13. Overview of R and D at TLK for process and analytical issues on tritium management in breeder blankets of ITER and DEMO

    International Nuclear Information System (INIS)

    Demange, D.; Alecu, C.G.; Bekris, N.; Borisevich, O.; Bornschein, B.; Fischer, S.; Gramlich, N.; Köllö, Z.; Le, T.L.; Michling, R.; Priester, F.; Röllig, M.; Schlösser, M.; Stämmler, S.; Sturm, M.; Wagner, R.; Welte, S.

    2012-01-01

    Highlights: ► We present advanced processes and analytics to improve tritium management. ► Membranes and membrane reactors can minimise tritium residence time and inventory. ► Spectroscopic methods can ensure on-line and near to real time tritium measurement. - Abstract: Safe, reliable, and efficient tritium management in the breeder blanket will have to face unprecedented technological challenges. Beside the efficiency for tritium recovery from the breeder blanket (Tritium Extraction (TES) and Coolant Purification Systems (CPS)), the accuracy for tritium tracking between the inner and the outer fuel cycle must also be demonstrated. This paper focuses on the recent R and D carried out at the Tritium Laboratory Karlsruhe to tackle these issues. For ITER, the recently consolidated TES and CPS designs comprise adsorption columns and getter beds operated in semi-continuous mode. Different approaches for the tritium accountancy stage (TAS) have been evaluated. Balancing static (batch-wise gas collection at the TBM outlets and the tritium plant) or dynamic (in/on-line) approaches with respect to the expected analytical performances and integration issues, the first conceptual design of the TAS for EU TBMs is presented. For DEMO, the overall strategy for tritium recovery and tracking has been revisited. The necessity for on-line real-time tritium accountancy and improved process efficiency suggest the use of continuous processes such as permeator and catalytic membrane reactor. The main benefits combining the PERMCAT process with advanced membranes is discussed with respect to process improvements and facilitated accountancy using spectroscopic methods.

  14. 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

  15. HHF test with 80x80x1 Be/Cu/SS Mock-ups for verifying the joining technology of the ITER blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Hong, Bong Guen; Jeong, Yong Hwan; Park, Jeong Yong; Choi, Byung Kwon; Jung, Hyun Kyu

    2008-11-15

    Through the fabrication of the Cu/SS and Be/Cu joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr for Cu/SS, 580 .deg. C 100 MPa 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, several HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS Mock-ups. In the present study, the HHF test with Be/Cu/SS Mock-ups, which have 80 mm x 80 mm single Be tile and each material depths were kept to be the same as the ITER blanket FW. The Mock-ups fabricated with three kinds of interlayers such as Cr/Ti/Cu, Ti/Cr/Cu, Ti/Cu, which were different from the developed interlayer (Cr/Cu), total 6 Mock-ups were fabricated. Preliminary analysis were performed to decide the test conditions; they were tested with up to 2.5 MW/m2 of heat fluxes and 20 cycles for each Mock-up in a given heat flux. They were tested with JUDITH-1 at FZJ in Germany. During tests, all Mock-ups showed delamination or full detachment of Be tile and it can be concluded that the joints with these interlayers have a bad joining but it can be used as a good data for developing the Be/Cu joint with HIP.

  16. RELAP/SCDAPSIM/MOD4.0 modification for transient accident scenario of Test Blanket Modules in ITER involving helium flows into heavy liquid metal

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J.; Pérez, M.; Mas de les Valls, E.; Batet, L.; Sandeep, T.; Chaudhari, V.; Reventós, F.

    2015-07-01

    The Institute for Plasma Research (IPR), India, is currently involved in the design and development of its Test Blanket Module (TBM) for testing in ITER (International Thermo nuclear Experimental Reactor). The Indian TBM concept is a Lead-Lithium cooled Ceramic Breeder (LLCB), which utilizes lead-lithium eutectic alloy (LLE) as tritium breeder, neutron multiplier and coolant. The first wall facing the plasma is cooled by helium gas. In preparation of the regulatory safety files of ITER-TBM, a number of off-normal event sequences have been postulated. Thermal hydraulic safety analyses of the TBM system will be carried out with the system code RELAP/SCDAPSIM/MOD4.0 which was initially designed to predict the behavior of light water reactor systems during normal and accidental conditions. In order to analyze some of the postulated off-normal events, there is the need to simulate the mixing of Helium and Lead-Lithium fluids. The Technical University of Catalonia is cooperating with IPR to implement the necessary changes in the code to allow for the mixing of helium and liquid metal. In the present study, the RELAP/SCDAPSIM/MOD4 two-phase flow 6-equations structure has been modified to allow for the mixture of LLE in the liquid phase with dry Helium in the gas phase. Practically obtaining a two-fluid 6-equation model where each fluid is simulated with a set of energy, mass and momentum balance equations. A preliminary flow regime map for LLE and helium flow has been developed on the basis of numerical simulations with the OpenFOAM CFD toolkit. The new code modifications have been verified for vertical and horizontal configurations. (Author)

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

    International Nuclear Information System (INIS)

    Stringer, J.; Blevins, J.

    1992-01-01

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

  18. 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

  19. Conceptual design on interface between ITER and tritium extraction system of Chinese helium-cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Zhang Long; Luo Tianyong; Feng Kaiming

    2010-01-01

    Tritium extraction system is essential for CN HCSB TBM for safety and technical reasons. Based on the assessments of system functions, integration issues and safety considerations, two main modifications of the system from previous design (Feng et al., 2007 ; Chen et al., 2008 ) are adopted: a)the TES has been split to 2 parts with one in port cell and another in tritium building. Q 2 O in the purge gas is reduced to Q 2 in a hot metal bed located in port cell; Q 2 is separated from the stream by a pair of cryogenic molecular sieve beds and a Pd/Ag diffuser located in tritium building. b)isotope separation process has been excluded. TES components sizes are estimated and space allocations are estimated. Required services and where and when they are needed are preliminary defined. Fluids delivered towards ITER tritium system are analyzed.

  20. An iterative approach for the optimization of pavement maintenance management at the network level.

    Science.gov (United States)

    Torres-Machí, Cristina; Chamorro, Alondra; Videla, Carlos; Pellicer, Eugenio; Yepes, Víctor

    2014-01-01

    Pavement maintenance is one of the major issues of public agencies. Insufficient investment or inefficient maintenance strategies lead to high economic expenses in the long term. Under budgetary restrictions, the optimal allocation of resources becomes a crucial aspect. Two traditional approaches (sequential and holistic) and four classes of optimization methods (selection based on ranking, mathematical optimization, near optimization, and other methods) have been applied to solve this problem. They vary in the number of alternatives considered and how the selection process is performed. Therefore, a previous understanding of the problem is mandatory to identify the most suitable approach and method for a particular network. This study aims to assist highway agencies, researchers, and practitioners on when and how to apply available methods based on a comparative analysis of the current state of the practice. Holistic approach tackles the problem considering the overall network condition, while the sequential approach is easier to implement and understand, but may lead to solutions far from optimal. Scenarios defining the suitability of these approaches are defined. Finally, an iterative approach gathering the advantages of traditional approaches is proposed and applied in a case study. The proposed approach considers the overall network condition in a simpler and more intuitive manner than the holistic approach.

  1. An Iterative Approach for the Optimization of Pavement Maintenance Management at the Network Level

    Directory of Open Access Journals (Sweden)

    Cristina Torres-Machí

    2014-01-01

    Full Text Available Pavement maintenance is one of the major issues of public agencies. Insufficient investment or inefficient maintenance strategies lead to high economic expenses in the long term. Under budgetary restrictions, the optimal allocation of resources becomes a crucial aspect. Two traditional approaches (sequential and holistic and four classes of optimization methods (selection based on ranking, mathematical optimization, near optimization, and other methods have been applied to solve this problem. They vary in the number of alternatives considered and how the selection process is performed. Therefore, a previous understanding of the problem is mandatory to identify the most suitable approach and method for a particular network. This study aims to assist highway agencies, researchers, and practitioners on when and how to apply available methods based on a comparative analysis of the current state of the practice. Holistic approach tackles the problem considering the overall network condition, while the sequential approach is easier to implement and understand, but may lead to solutions far from optimal. Scenarios defining the suitability of these approaches are defined. Finally, an iterative approach gathering the advantages of traditional approaches is proposed and applied in a case study. The proposed approach considers the overall network condition in a simpler and more intuitive manner than the holistic approach.

  2. Tritium breeding blanket

    International Nuclear Information System (INIS)

    Smith, D.; Billone, M.; Gohar, Y.; Baker, C.; Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H.; Raffray, A.; Sviatoslavsky, I.; Simbolotti, G.; Shatalov, G.

    1991-01-01

    The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R ampersand D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs

  3. 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.

  4. Introducing artificial depth cues to improve task performance in ITER maintenance actions

    International Nuclear Information System (INIS)

    Heemskerk, C.J.M.; Eendebak, P.T.; Schropp, G.Y.R.; Hermes, H.V.; Elzendoorn, B.S.Q.; Magielsen, A.J.

    2013-01-01

    Highlights: ► Operators regard lack of 3D perception as primary factor hindering remote maintenance. ► Promising techniques to improve depth perception are depth gauges and stereo vision. ► First experiment shows artificial depth gauges need to be designed with care. ► Second experiment shows that stereo vision improves task performance significantly. -- Abstract: Maintenance operations on ITER tokamak components will be largely performed by remote handling. In previous work it was shown that representative maintenance tasks could be performed significantly faster with direct visual feedback than with camera feedback. In post-test interviews, operators indicated that they regarded the lack of 3D perception as the primary factor hindering their performance. This paper discusses various techniques to improve depth perception in teleoperation, including stereo vision, head tracking, virtual camera views and depth gauges. The most promising techniques were tested. Performance metrics included time-to-complete, path analysis and operator work-load. In a first experiment, artificial depth gauges views were tested in a 1:1 scale hardware testbed with mechanical master-slave manipulators handled by experienced operators. Robust real-time image processing was achieved with marker-based objects. The simple depth gauge and graphical overlay did not significantly improve task performance. Operators commented on their view of the task being “obstructed” by the graphical overlay, and the depth gauge was judged not very informative. In a second experiment, real time tracking was combined with VR display including stereo and head tracking. While stereo was found to improve the task performance significantly over the 1 camera (mono) condition, head tracking unexpectedly did not

  5. Introducing artificial depth cues to improve task performance in ITER maintenance actions

    Energy Technology Data Exchange (ETDEWEB)

    Heemskerk, C.J.M., E-mail: c.heemskerk@heemskerk-innovative.nl [Heemskerk Innovative Technology, Sassenheim (Netherlands); Eendebak, P.T. [TNO Netherlands Organisation for Applied Scientific Research (Netherlands); Schropp, G.Y.R. [Heemskerk Innovative Technology, Sassenheim (Netherlands); Hermes, H.V. [TU Eindhoven (Netherlands); Elzendoorn, B.S.Q. [FOM Institute DIFFER (Dutch Institute for Fundamental Energy Research), Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Magielsen, A.J. [NRG, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2013-10-15

    Highlights: ► Operators regard lack of 3D perception as primary factor hindering remote maintenance. ► Promising techniques to improve depth perception are depth gauges and stereo vision. ► First experiment shows artificial depth gauges need to be designed with care. ► Second experiment shows that stereo vision improves task performance significantly. -- Abstract: Maintenance operations on ITER tokamak components will be largely performed by remote handling. In previous work it was shown that representative maintenance tasks could be performed significantly faster with direct visual feedback than with camera feedback. In post-test interviews, operators indicated that they regarded the lack of 3D perception as the primary factor hindering their performance. This paper discusses various techniques to improve depth perception in teleoperation, including stereo vision, head tracking, virtual camera views and depth gauges. The most promising techniques were tested. Performance metrics included time-to-complete, path analysis and operator work-load. In a first experiment, artificial depth gauges views were tested in a 1:1 scale hardware testbed with mechanical master-slave manipulators handled by experienced operators. Robust real-time image processing was achieved with marker-based objects. The simple depth gauge and graphical overlay did not significantly improve task performance. Operators commented on their view of the task being “obstructed” by the graphical overlay, and the depth gauge was judged not very informative. In a second experiment, real time tracking was combined with VR display including stereo and head tracking. While stereo was found to improve the task performance significantly over the 1 camera (mono) condition, head tracking unexpectedly did not.

  6. Intelligent controller of a flexible hybrid robot machine for ITER assembly and maintenance

    International Nuclear Information System (INIS)

    Al-saedi, Mazin I.; Wu, Huapeng; Handroos, Heikki

    2014-01-01

    Highlights: • Studying flexible multibody dynamic of hybrid parallel robot. • Investigating fuzzy-PD controller to control a hybrid flexible hydraulically driven robot. • Investigating ANFIS-PD controller to control a hybrid flexible robot. Compare to traditional PID this method gives better performance. • Using the equilibrium of reaction forces between the parallel and serial parts of hybrid robot to control the serial part hydraulically driven. - Abstract: The assembly and maintenance of International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is highly challenging since the tasks performed by the robot involve welding, material handling, and machine cutting from inside the VV. To fulfill the tasks in ITER application, this paper presents a hybrid redundant manipulator with four DOFs provided by serial kinematic axes and six DOFs by parallel mechanism. Thus, in machining, to achieve greater end-effector trajectory tracking accuracy for surface quality, a robust control of the actuators for the flexible link has to be deduced. In this paper, the intelligent control of a hydraulically driven parallel robot part based on the dynamic model and two control schemes have been investigated: (1) fuzzy-PID self tuning controller composed of the conventional PID control and with fuzzy logic; (2) adaptive neuro-fuzzy inference system-PID (ANFIS-PID) self tuning of the gains of the PID controller, which are implemented independently to control each hydraulic cylinder of the parallel robot based on rod position predictions. The obtained results of the fuzzy-PID and ANFIS-PID self tuning controller can reduce more tracking errors than the conventional PID controller. Subsequently, the serial component of the hybrid robot can be analyzed using the equilibrium of reaction forces at the universal joint connections of the hexa-element. To achieve precise positional control of the end effector for maximum precision machining, the hydraulic cylinder should

  7. Intelligent controller of a flexible hybrid robot machine for ITER assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Al-saedi, Mazin I., E-mail: mazin.al-saedi@lut.fi; Wu, Huapeng; Handroos, Heikki

    2014-10-15

    Highlights: • Studying flexible multibody dynamic of hybrid parallel robot. • Investigating fuzzy-PD controller to control a hybrid flexible hydraulically driven robot. • Investigating ANFIS-PD controller to control a hybrid flexible robot. Compare to traditional PID this method gives better performance. • Using the equilibrium of reaction forces between the parallel and serial parts of hybrid robot to control the serial part hydraulically driven. - Abstract: The assembly and maintenance of International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is highly challenging since the tasks performed by the robot involve welding, material handling, and machine cutting from inside the VV. To fulfill the tasks in ITER application, this paper presents a hybrid redundant manipulator with four DOFs provided by serial kinematic axes and six DOFs by parallel mechanism. Thus, in machining, to achieve greater end-effector trajectory tracking accuracy for surface quality, a robust control of the actuators for the flexible link has to be deduced. In this paper, the intelligent control of a hydraulically driven parallel robot part based on the dynamic model and two control schemes have been investigated: (1) fuzzy-PID self tuning controller composed of the conventional PID control and with fuzzy logic; (2) adaptive neuro-fuzzy inference system-PID (ANFIS-PID) self tuning of the gains of the PID controller, which are implemented independently to control each hydraulic cylinder of the parallel robot based on rod position predictions. The obtained results of the fuzzy-PID and ANFIS-PID self tuning controller can reduce more tracking errors than the conventional PID controller. Subsequently, the serial component of the hybrid robot can be analyzed using the equilibrium of reaction forces at the universal joint connections of the hexa-element. To achieve precise positional control of the end effector for maximum precision machining, the hydraulic cylinder should

  8. Status of reliability in determining SDDR for manual maintenance activities in ITER: Quality assessment of relevant activation cross sections involved

    International Nuclear Information System (INIS)

    Garcia, R.; Garcia, M.; Pampin, R.; Sanz, J.

    2016-01-01

    Highlights: • Feasibility of manual maintenance activities in ITER port cell and port interspace. • Activation of relevant materials and components placed in the current ITER model. • Dominant radionuclides and pathways for shutdown dose rate in ITER. • Quality analysis of typically used EAF and TENDL activation libraries is performed. • EAF performance found as trustworthy with slight recommended improvements. - Abstract: This paper assesses the quality of the EAF-2007 and 2010 activation cross sections for relevant reactions in the determination of the Shutdown Dose Rate (SDDR) in the Port Cell (PC) and Port Interspace (PI) areas of ITER. For each of relevant ITER materials, dominant radionuclides responsible of SDDR and their production pathways are listed. This information comes from a review of the recent reports/papers about SDDR in ITER and own calculations. A total of 26 relevant pathways are found. The quality of these cross sections pathways is assessed following EAF validation procedure, and for those found as not validated last TENDL library versions have been investigated in order to check possible improvements when compared to EAF. The use of EAF libraries is found as trustworthy and it is recommended for the prediction of SDDR in the ITER PC and PI. However, 3 cross section reactions are considered for further improvement: Co59(n,2n)Co58, Cu63(n,g)Cu64 and Cr50(n,g)Cr51.

  9. Status of reliability in determining SDDR for manual maintenance activities in ITER: Quality assessment of relevant activation cross sections involved

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, R., E-mail: rgarciam@ind.uned.es [UNED, Power Engineering Department, C/Juan del Rosal 12, 28040 Madrid (Spain); Garcia, M. [UNED, Power Engineering Department, C/Juan del Rosal 12, 28040 Madrid (Spain); Pampin, R. [F4E, Torres Diagonal Litoral B3, Barcelona (Spain); Sanz, J. [UNED, Power Engineering Department, C/Juan del Rosal 12, 28040 Madrid (Spain)

    2016-11-15

    Highlights: • Feasibility of manual maintenance activities in ITER port cell and port interspace. • Activation of relevant materials and components placed in the current ITER model. • Dominant radionuclides and pathways for shutdown dose rate in ITER. • Quality analysis of typically used EAF and TENDL activation libraries is performed. • EAF performance found as trustworthy with slight recommended improvements. - Abstract: This paper assesses the quality of the EAF-2007 and 2010 activation cross sections for relevant reactions in the determination of the Shutdown Dose Rate (SDDR) in the Port Cell (PC) and Port Interspace (PI) areas of ITER. For each of relevant ITER materials, dominant radionuclides responsible of SDDR and their production pathways are listed. This information comes from a review of the recent reports/papers about SDDR in ITER and own calculations. A total of 26 relevant pathways are found. The quality of these cross sections pathways is assessed following EAF validation procedure, and for those found as not validated last TENDL library versions have been investigated in order to check possible improvements when compared to EAF. The use of EAF libraries is found as trustworthy and it is recommended for the prediction of SDDR in the ITER PC and PI. However, 3 cross section reactions are considered for further improvement: Co59(n,2n)Co58, Cu63(n,g)Cu64 and Cr50(n,g)Cr51.

  10. Development of filler wires for welding of reduced activation ferritic martensitic steel for India's test blanket module of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2010-07-01

    Indigenous development of reduced activation ferritic-martensitic (RAFM) steel has become necessary for India as a participant in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFM steel is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFM steel filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFM steel. The purpose of this study is to develop filler wires that can be directly used for both gas tungsten arc welding (GTAW) and for narrow-gap gas tungsten arc welding (NG-GTAW) that reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser-MIG welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using GTAW process at various heat inputs with a preheat temperature of 250 C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some amount of delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimized to qualify the filler wires without the presence of delta-ferrite in the weld metal and with optimized mechanical properties. Results showed that the weld metals are free from delta-ferrite. Tensile properties at ambient temperature and at 500 C are well above the specified values, and are much higher than the base metal values. Ductile Brittle Transition Temperature (DBTT) has been evaluated as -81 C based on the 68 J criteria. The present study highlights the basis and methodology

  11. Development of filler wires for welding of reduced activation ferritic martenstic steel for India's test blanket module of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, G., E-mail: gsrini@igcar.gov.in [Materials Technology Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamilnadu (India); Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K. [Materials Technology Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamilnadu (India)

    2011-06-15

    Highlights: > Weld microstructure produced by RAFMS filler wires are free from delta ferrite. > Cooling rates of by weld thermal cycles influences the presence of delta ferrite. > Weld parameters modified with higher pre heat temperature and high heat input. > PWHT optimized based on correlation of hardness between base and weld metals. > Optimised mechanical properties achieved by proper tempering of the martensite. - Abstract: Indigenous development of reduced activation ferritic martensitic steel (RAFMS) has become mandatory to India to participate in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFMS is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFMS filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFMS. Purpose of this study is to develop filler wires that can be directly used for both tungsten inert gas welding (TIG) and narrow gap tungsten inert gas welding (NG-TIG), which reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, autogenous welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using TIG process at various heat inputs with a preheat temperature of 250 deg. C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimised to qualify the filler wires without the presence of delta-ferrite in the weld

  12. Design and Analysis of the Korean Small Semi-prototype Mock-up for the 2nd Qualification of the ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Yoon, Jae Sung; Lee, Eo Hwak; Lee, Seung Jae; Choi, Bo Guen; Park, Jeong Yong; Jung, Yang Il; Choi, Byung Kwon; Kim, Byoung Yoon

    2011-01-01

    Since the blanket First Wall (FW) of the International Thermonuclear Experimental Reactor (ITER) is subjected to a high heat and high neutron loads, it is one of the most important components. It composed of a beryllium (Be) layer as a plasma facing material, a copper alloy (CuCrZr) layer as a heat sink and type 316L authentic stainless steel (SS316L) as a structure material. The joining of the three different metals is the key issue to be solved. And more, the peak heat load was assumed to be 0.5 MW/m 2 in the initial design of the FW, but it was changed to be up to 5 MW/m 2 In Korea, the joining method has developed and it was proved through the several mock-up fabrication and high heat flux tests for confirming the joining integrity. Some of them were tested in the foreign facilities such as JEBIS at JAEA in Japan, TSEFEY at Efremov in Russia, and JUDITH at FZJ in Germany, and others were tested in our own facilities such as KoHLT-1 and -2. And finally, the 1 st , recently. Therefore, the FW panel design has been changed for enhancing the cooling and ITER Organization will provide the proposed design. Qualification was passed, in which two 80x80x3 Be/Cu/SS mock-ups were tested under 0.625 and 0.875 MW/m 2 heat fluxes for 12,000 cycles and then tested under 1.75 and 1.40 MW/m 2 Currently, the 2 heat fluxes for 1,000 cycles at FZJ and SNL, respectively. Currently, the 2 nd qualification program was started and the semi-prototype should be fabricated by the end of 2011 for testing under 5.0 MW/m 2 heat flux for certain number of cycles. In order to prepare the semi-prototype, several fabrication methods should be developed through the fabrication and test with the several mock-ups. In the present study, small Be mock-up was fabricated as the first step for the preparation. It was fabricated according to the designs considering the currently modified design of the FW. In the present paper, the fabrication objectives, methods, results and related tests were

  13. Development of a Remote Handling Robot for the Maintenance of an ITER-Like D-Shaped Vessel

    Directory of Open Access Journals (Sweden)

    Peihua Chen

    2014-01-01

    Full Text Available Robotic operation is one of the major challenges in the remote maintenance of ITER vacuum vessel (VV and future fusion reactors as inner operations of Tokamak have to be done by robots due to the internal adverse conditions. This paper introduces a novel remote handling robot (RHR for the maintenance of ITER-like D-shaped vessel. The modular designed RHR, which is an important part of the remote handling system for ITER, consists of three parts: an omnidirectional transfer vehicle (OTV, a planar articulated arm (PAA, and an articulated teleoperated manipulator (ATM. The task of RHR is to carry processing tools, such as the viewing system, leakage detector, and electric screwdriver, to inspect and maintain the components installed inside the D-shaped vessel. The kinematics of the OTV, as well as the kinematic analyses of the PAA and ATM, is studied in this paper. Because of its special length and heavy payload, the dynamics of the PAA is also investigated through a dynamic simulation system based on robot technology middleware (RTM. The results of the path planning, workspace simulations, and dynamic simulation indicate that the RHR has good mobility together with satisfying kinematic and dynamic performances and can well accomplish its maintenance tasks in the ITER-like D-shaped vessel.

  14. New achievements of the Divertor Test Platform programme for the ITER divertor remote maintenance R and D

    International Nuclear Information System (INIS)

    Damiani, C.; Baldi, L.; Galbiati, L.; Irving, M.; Lorenzelli, L.; Micciche, G.; Muro, L.; Nucci, S.; Varocchi, G.; Poggianti, A.; Fermani, G.; Maisonnier, D.; Palmer, J.; Martin, E.; Friconneau, J.P.; Gravez, P.; Takeda, N.

    2001-01-01

    The divertor assembly for the ITER fusion reactor consists of a number of rail mounted cassettes (54 now in ITER FEAT) located in the bottom region of the vacuum vessel. These cassettes shall be removed/installed remotely during the life of the reactor by means of specific devices. To demonstrate and optimise the feasibility of the in-vessel maintenance process the Divertor Test Platform (DTP) has been established at the ENEA Research Centre in Brasimone, Italy, as a major part of the large ITER R and D project L7. A first set of tests has been already carried out and reported during 1998, when the basic feasibility of the divertor replacement was demonstrated. In the present period (January 1999-July 2000), new activities, including both site tests and other 'external' R and D works, have been carried out in order to refine and improve the ITER divertor maintenance scenario. These include the study of abnormal maintenance operations and of possible handling equipment failure and its consequences; the procurement and testing of new sub-systems (e.g. a force reflection manipulator arm), and the development of remote handling techniques including a virtual reality system. Following a short description of the DTP, this paper reports on the new results and achievements, draws the relevant conclusions, and finally discusses future activities

  15. Design of an overhead crane for the ITER NB cell remote handling maintenance operations

    Energy Technology Data Exchange (ETDEWEB)

    Taubmann, Gonzalo; Brochet, Laurent [IBERTEF A.I.E., Iberica de Tecnologia de Fusion, C/Magallanes 3, 28015 Madrid (Spain); Liniers, Macarena [Asociacion EURATOM-CIEMAT para la Fusion, Av. Complutense 22, 28040 Madrid (Spain)], E-mail: macarena.liniers@ciemat.es; Medrano, Mercedes; Sarasola, Xabier; Botija, Jose; Alonso, Javier [Asociacion EURATOM-CIEMAT para la Fusion, Av. Complutense 22, 28040 Madrid (Spain); Damiani, Carlo [FUSION FOR ENERGY, Josep Pla 2, Torres Diagonal Litoral Ed B3, 08019 Barcelona (Spain)

    2009-06-15

    In the neutral beam cell of ITER all the maintenance operations on the neutral beam components (BLC's) must be performed by an overhead crane of large payload capability (30-50 tonnes). A crane system is presented consisting of a monorail, a carriage, and a lifting mechanism. The monorail must give access to the BLC's in the beam line vessel, the front components connecting the NB vessel with the Tokamak, and a storage area at the north end of the NB cell. Rail switching points are required at the intersections between radial and toroidal branches. A translational switching mechanism is proposed. The crane carriage consists of two independent sub-carriages, each composed of four wheels. A set of four secondary wheels attached to the main carriage prevents the crane tilting due to the CoG misalignment of some loads. The elevation system proposed consists of an electromechanical crane of four independent drums and 50 tonnes payload. In parallel with the crane design, a logistics and space availability study has been carried out, leading to the detection of clearance or transport problems that could be taken into account in the final crane design.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  18. Design of an overhead crane for the ITER NB cell remote handling maintenance operations

    International Nuclear Information System (INIS)

    Taubmann, Gonzalo; Brochet, Laurent; Liniers, Macarena; Medrano, Mercedes; Sarasola, Xabier; Botija, Jose; Alonso, Javier; Damiani, Carlo

    2009-01-01

    In the neutral beam cell of ITER all the maintenance operations on the neutral beam components (BLC's) must be performed by an overhead crane of large payload capability (30-50 tonnes). A crane system is presented consisting of a monorail, a carriage, and a lifting mechanism. The monorail must give access to the BLC's in the beam line vessel, the front components connecting the NB vessel with the Tokamak, and a storage area at the north end of the NB cell. Rail switching points are required at the intersections between radial and toroidal branches. A translational switching mechanism is proposed. The crane carriage consists of two independent sub-carriages, each composed of four wheels. A set of four secondary wheels attached to the main carriage prevents the crane tilting due to the CoG misalignment of some loads. The elevation system proposed consists of an electromechanical crane of four independent drums and 50 tonnes payload. In parallel with the crane design, a logistics and space availability study has been carried out, leading to the detection of clearance or transport problems that could be taken into account in the final crane design.

  19. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

    In fusion reactors, the blanket design and its characteristics have a major impact on the reactor performance, size, and economics. The selection and arrangement of the blanket materials, dimensions of the different blanket zones, and different requirements of the selected materials for a satisfactory performance are the main parameters, which define the blanket performance. These parameters translate to a large number of variables and design constraints, which need to be simultaneously considered in the blanket design process. This represents a major design challenge because of the lack of a comprehensive design tool capable of considering all these variables to define the optimum blanket design and satisfying all the design constraints for the adopted figure of merit and the blanket design criteria. The blanket design techniques of the First Wall/Blanket/Shield Design and Optimization System (BSDOS) have been developed to overcome this difficulty and to provide the state-of-the-art techniques and tools for performing blanket design and analysis. This report describes some of the BSDOS techniques and demonstrates its use. In addition, the use of the optimization technique of the BSDOS can result in a significant blanket performance enhancement and cost saving for the reactor design under consideration. In this report, examples are presented, which utilize an earlier version of the ITER solid breeder blanket design and a high power density self-cooled lithium blanket design for demonstrating some of the BSDOS blanket design techniques

  20. Disruption problematics in segmented blanket concepts

    International Nuclear Information System (INIS)

    Crutzen, Y.; Fantechi, S.; Farfaletti-Casali, F.

    1994-01-01

    In Tokamaks, the hostile operating environment originated by plasma disruption events requires that the first wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence there is a need to improve the safety features of the blanket design concepts satisfying the disruption problematics and to formulate guidelines on the required internal reinforcements of the blanket components. The present paper describes the recent investigations on blanket reinforcement systems needed in order to optimize the first-wall/blanket/shield structural design for next step and commercial fusion reactors in the context of ITER, DEMO and SEAFP activities

  1. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

    2006-02-01

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17LI is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Ph-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure.

  2. Initial three-dimensional neutronics calculations for the EU water cooled lithium-lead test blanket module for ITER-FEAT

    International Nuclear Information System (INIS)

    Jordanova, J.; Poitevin, Y.; Li Puma, A.; Kirov, N.

    2003-01-01

    The paper summarizes the main results of the initial three-dimensional radiation transport analysis of the EU water-cooled lithium-lead test blanket module performed using the Monte Carlo code MCNP. Estimates of tritium production rate, nuclear energy deposition and cumulative fluence effects such as radiation damage through atomic displacement and production of He and H are presented. (author)

  3. Mechanical design issues associated with mounting, maintenance, and handling of an ITER divertor

    International Nuclear Information System (INIS)

    Goranson, D.L.; Fogarty, D.J.; Jones, G.H.

    1992-01-01

    Several designs that address plasma-facing plate configurations and thermal-hydraulic design issues have been developed for the ITER divertor. Design criteria growing out of physics requirements, physical constraints, and remote handling requirements impose severe mechanical requirements on the support structure and its attachments. These pose a challenge to the mechanical design of a divertor, which must be addressed before a functional divertor is practical that is, one that can be remotely handled, aligned, and maintained; that functions reliably under thermal loading and disruptions; and that gives the required life in the nuclear environment predicted for ITER. This paper discusses the design criteria for the divertor mounting structure and identifies the mechanical design issues that need to be addressed

  4. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

    Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-10-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

  5. Beryllium R and D for blanket application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

    1998-01-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

  6. Beryllium R&D for blanket application

    Science.gov (United States)

    Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

    1998-10-01

    The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

  7. Magnetoconvection in HCLL blankets

    International Nuclear Information System (INIS)

    Mistrangelo, C.; Buehler, L.

    2014-01-01

    In the present work we consider magneto-convective flows in one of the proposed European liquid metal blankets that will be tested in the experimental fusion reactor ITER. Here the PbLi alloy is used as breeder material and helium as coolant. In order to finalize the design of the helium cooled lead lithium (HCLL) blanket, studies are still required to fully understand the behavior of the electrically conducting breeder under the influence of the intense magnetic field that confines the fusion plasma and in case of non-uniform thermal conditions. Liquid metal HCLL blanket flows are expected to be mainly driven by buoyancy forces caused by non-isothermal operating conditions due to neutron volumetric heating and cooling of walls, since only a weak forced ow is foreseen for tritium extraction in external ancillary systems. Buoyancy can therefore become very important and modify the velocity distribution and related heat transfer performance of the blanket. The present numerical study aims at clarifying the influence of electromagnetic and thermal coupling of neighboring fluid domains on magneto-convective flows in geometries relevant for the HCLL blanket concept. According to the last design review two internal cooling plates subdivide the fluid domain into three slender flow regions, which are thermally and electrically coupled through common walls. First a uniform volumetric heat source is considered to identify the basic convective patterns that establish in the liquid metal. Results are then compared with those obtained by applying a realistic radial distribution of the power density as obtained from a neutronic analysis. Velocity and temperature distributions are discussed for various volumetric heat sources and magnetic field strengths.

  8. Design and implementation of motion planning of inspection and maintenance robot for ITER-like vessel

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hesheng; Lai, Yinping [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Cao, Qixin [Institute of Robotics, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2015-12-15

    Robot motion planning is a fundamental problem to ensure the robot executing the task without clashes, fast and accurately in a special environment. In this paper, a motion planning of a 12 DOFs remote handling robot used for inspecting the working state of the ITER-like vessel and maintaining key device components is proposed and implemented. Firstly, the forward and inverse kinematics are given by analytic method. The work space and posture space of this manipulator are both considered. Then the motion planning is divided into three stages: coming out of the cassette mover, moving along the in-vessel center line, and inspecting the D-shape section. Lastly, the result of experiments verified the performance of the motion design method. In addition, the task of unscrewing/screwing the screw demonstrated the feasibility of system in function.

  9. Test Blanket Working Group's recent activities

    International Nuclear Information System (INIS)

    Vetter, J.E.

    2001-01-01

    The ITER Test Blanket Working Group (TBWG) has continued its activities during the period of extension of the EDA with a revised charter on the co-ordination of the development work performed by the Parties and by the JCT leading to a co-ordinated test programme on ITER for a DEMO-relevant tritium breeding blanket. This follows earlier work carried out until July 1998, which formed part of the ITER Final Design Report (FDR), completed in 1998. Whilst the machine parameters for ITER-FEAT have been significantly revised compared to the FDR, testing of breeding blanket modules remains a main objective of the test programme and the development of a reactor-relevant breeding blanket to ensure tritium fuel self-sufficiency is recognized a key issue for fusion. Design work and R and D on breeding blanket concepts, including co-operation with the other Contacting Parties of the ITER-EDA for testing these concepts in ITER, are included in the work plans of the Parties

  10. Development of radiation-hard electric connector with ball bearing for in-vessel remote maintenance equipment of ITER

    International Nuclear Information System (INIS)

    Ito, Akira; Obara, Kenjiro; Tada, Eisuke; Morita, Yousuke; Yagi, Toshiaki; Iida, Kazuhisa; Sato, Masaru.

    1997-12-01

    Development of radiation-hard electric connector with ball bearing for in-vessel remote maintenance equipment of ITER (International Thermonuclear Experimental Reactor) has been conducted. Since the in-vessel remote maintenance equipment is operated under the condition of 10 6 R/h gamma ray dose rate, the electric connector has to be radiation hard for an accumulation dose of 10 10 R. In addition, the simple attachment/removal mechanism is essential for remote operation. For this, the alumina (Al203) ceramics and a ball bearing were adopted to electric insulator and plug (male) of connector, respectively. The handling tests on attachment/removal of the connector were conducted by using master slave manipulator and general purpose robot with handling tool, and as a result, the validity of the attachment/removal mechanism was verified. In the gamma ray irradiation tests, which are under way, no degradation in break down voltage (1000V 1min.) up to 10 10 R was confirmed. However insulation resistance and contact resistance between contact pin and contact socket were deteriorated in proportion to the accumulation dose. Increase of contact resistance is considered due to an erosion of contact pin. (author)

  11. Conceptual design of Blanket Remote Handling System for CFETR

    International Nuclear Information System (INIS)

    Wei, Jianghua; Song, Yuntao; Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong

    2015-01-01

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  12. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2015-11-15

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  13. A PRELIMINARY ASSESSMENT OF THE OCCUPATIONAL RADIATION EXPOSURE FROM MAINTAINING THE US ITER DCLL TBM

    Energy Technology Data Exchange (ETDEWEB)

    B. J. Merrill; L. C. Cadwallader; M. Dagher

    2008-09-01

    This paper details an Occupational Radiation Exposure (ORE) analysis performed for the US International Thermonuclear Experimental Reactor (ITER) Dual Coolant Lead Lithium (DCLL) Test Blanket Module (TBM). This ORE analysis was performed with the QADMOD dose code for maintenance activities anticipated for the US DCLL TBM concept and its ancillary systems. Identification of the maintenance tasks that will have to be performed and estimates of the time required to perform these tasks were developed based on either expert opinion or on industrial maintenance experience for similar technologies. This paper details the modeling activity and the calculated doses for the maintenance activities envisioned for the US DCLL TBM.

  14. Helium Loop for the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Neuberger, H.; Boccaccini, L.V.; Ghidersa, B. E.; Jin, X.; Meyder, R.

    2006-01-01

    In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium loop for the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) in ITER has been investigated with regard to the layout definition, selection of components, control, dimensioning and integration. This paper presents the status of development. The loop design for the HCPB-TBM in ITER will mainly base on the experience gained from Helium Loop Karlsruhe (HELOKA) which is currently developed at the FZK for experiments under ITER relevant conditions. The ITER loop will be equipped with similar components like HELOKA and will mainly consist of a circulator with variable speed drive, a recuperator, an electric heater, a cooler, a dust filter and auxilary components e.g. pipework and valves. A Coolant Purification System (CPS) and a Pressure Control System (PCS) are foreseen to meet the requirements on coolant conditioning. To prepare a TBM for a new experimental campaign, a succession of operational states like '' cold maintenance '', '' baking '' and '' cold standby '' is required. Before a pulse operation, a '' hot stand-by '' state should be achieved providing the TBM with inlet coolant at nominal conditions. This operation modus is continued in the dwell time waiting for the successive pulse. A '' tritium out-gassing '' will be also required after several TBM-campaigns to remove the inventory rest of T in the beds for measurement purpose. The dynamic circuit behaviour during pulses, transition between different operational states as well as the behaviour in accident situations are investigated with RELAP. The main components of the loop will be accommodated inside the Tokamak Cooling Water System(TCWS)- vault from where the pipes require connection to the TBM which is attached to port 16 of the vacuum vessel. Therefore pipes across the ITER- building of about 110 m in length (each) are required. Additional equipment is also located in the port cell

  15. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

    General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

  16. Use of nuclear data sensitivity and uncertainty analysis for the design preparation of the HCLL breeder blanket mock-up experiment for ITER

    International Nuclear Information System (INIS)

    Kodeli, I.

    2007-01-01

    An experiment on a mock-up of the Test Blanket module based on Helium Cooled Lithium Lead (HCLL) concept will be performed in 2007 in the FNG utility in Frascati in order to study neutronics characteristics of the module and the performance of the computational tools in the accurate prediction of the neutron transport. With the objective to prepare and optimise the design of the mock-up in the sense to provide maximum information on the state-of-the-art of the cross section data the mock-up was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR), their sensitivity to the underlying basic cross sections, as well as the corresponding uncertainty estimations were calculated using the deterministic transport codes (DOORS package), the sensitivity/uncertainty code package SUSD3D and the VITAMIN-J/COVA covariance matrix libraries. The cross section reactions with largest contribution to the uncertainty in the calculation of the TPR were identified to be (n,2n) and (n,3n) reactions on plumb. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross sections. (author)

  17. Use of Nuclear Data Sensitivity and Uncertainty Analysis for the Design Preparation of the HCLL Breeder Blanket Mockup Experiment for ITER

    Directory of Open Access Journals (Sweden)

    I. Kodeli

    2008-01-01

    Full Text Available An experiment on a mockup of the test blanket module based on helium-cooled lithium lead (HCLL concept will be performed in 2008 in the Frascati Neutron Generator (FNG in order to study neutronics characteristics of the module and the accuracy of the computational tools. With the objective to prepare and optimise the design of the mockup in the sense to provide maximum information on the state-of-the-art of the cross-section data the mockup was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR, their sensitivity to the underlying basic cross-sections, as well as the corresponding uncertainties were calculated using the deterministic transport codes (DOORS package, the sensitivity/uncertainty code package SUSD3D, and the VITAMINJ/ COVA covariance matrix libraries. The cross-section reactions with largest contribution to the uncertainty of the calculated TPR were identified to be (n,2n and (n,3n reactions on lead. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross-sections.

  18. Remote maintenance for fusion experimental reactor

    International Nuclear Information System (INIS)

    Koizumi, Koichi; Takeda, Nobukazu

    2000-01-01

    Here was introduced on maintenance of reactor core portion operated by remote control among maintenance of the International Thermonuclear Experimental Reactor (ITER) begun on its design since 1988 under international cooperation of U.S.A., Europe, Russia and Japan. Every appliances constructing the reactor core portion is necessary to carry out all of their inspection and maintenance by using remote controlled apparatus because of their radiation due to neutron generated by DT combustion of plasma. For engineering design activity (EDA) in ITER, not only design and development of the remote control appliances but also design under consideration of remote maintenance for from structural design of maintained objective appliances to access method to appliances, transportation and preservation method of radiated matters, and out-reactor maintenance in a hot cell, is now under progress. Here were also reported on basic concept on maintenance and conservation of ITER, maintenance design of diverter and blanket with high maintenance frequency and present state on development of maintenance appliances. (G.K.)

  19. RAMI analysis for DEMO HCPB blanket concept cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

    2015-10-15

    Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

  20. Development of thick wall welding and cutting tools for ITER

    International Nuclear Information System (INIS)

    Nakahira, Masataka; Takahashi, Hiroyuki; Akou, Kentaro; Koizumi, Koichi

    1998-01-01

    The Vacuum Vessel, which is a core component of International Thermonuclear Experimental Reactor (ITER), is required to be exchanged remotely in a case of accident such as superconducting coil failure. The in-vessel components such as blanket and divertor are planned to be exchanged or fixed. In these exchange or maintenance operations, the thick wall welding and cutting are inevitable and remote handling tools are necessary. The thick wall welding and cutting tools for blanket are under developing in the ITER R and D program. The design requirement is to weld or cut the stainless steel of 70 mm thickness in the narrow space. Tungsten inert gas (TIG) arc welding, plasma cutting and iodine laser welding/cutting are selected as primary option. Element welding and cutting tests, design of small tools to satisfy space requirement, test fabrication and performance tests were performed. This paper reports the tool design and overview of welding and cutting tests. (author)

  1. Development of thick wall welding and cutting tools for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, Masataka; Takahashi, Hiroyuki; Akou, Kentaro; Koizumi, Koichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The Vacuum Vessel, which is a core component of International Thermonuclear Experimental Reactor (ITER), is required to be exchanged remotely in a case of accident such as superconducting coil failure. The in-vessel components such as blanket and divertor are planned to be exchanged or fixed. In these exchange or maintenance operations, the thick wall welding and cutting are inevitable and remote handling tools are necessary. The thick wall welding and cutting tools for blanket are under developing in the ITER R and D program. The design requirement is to weld or cut the stainless steel of 70 mm thickness in the narrow space. Tungsten inert gas (TIG) arc welding, plasma cutting and iodine laser welding/cutting are selected as primary option. Element welding and cutting tests, design of small tools to satisfy space requirement, test fabrication and performance tests were performed. This paper reports the tool design and overview of welding and cutting tests. (author)

  2. Convertible shielding to ceramic breeding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Kurasawa, Toshimasa; Sato, Satoshi; Nakahira, Masataka; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1995-05-01

    Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

  3. Remote assembly and maintenance of fusion reactors

    International Nuclear Information System (INIS)

    Becquet, M.C.; Farfaletti-Casali, F.

    1991-01-01

    This paper intend to present the state of the art in the field of remote assembly and maintenance, including system analysis design and operation for controlled fusion device such as JET, and the next NET and ITER reactors. The operational constraints of fusion reactors with respect to temperature, radiations dose rates and cumulated doses are considered with the resulting design requirements. Concepts like articulated boom, in-vessel vehicle and blanket handling device are presented. The close relations between computer simulations and experimental validation of those concepts are emphasized to ensure reliability of the operational behavior. Mockups and prototypes in reduced and full scale, as operating machines are described to illustrate the progress in remote operations for fusion reactors. The developments achieved at the Institute for System Engineering and Informatics of the Joint Research Center, in the field of remote blanket maintenance, reliability assessment of RH systems and remote cut and welding of lips joints are considered. (author)

  4. ITER EDA newsletter. V. 4, no. 9

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the first meeting of the ITER Test Blanket Working Group held 19-21 July 1995 at the ITER Garching Joint Work Site, and on the second workshop of the ITER Expert Group on Confinement and Transport.

  5. ITER EDA newsletter. V. 4, no. 9

    International Nuclear Information System (INIS)

    1995-09-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the first meeting of the ITER Test Blanket Working Group held 19-21 July 1995 at the ITER Garching Joint Work Site, and on the second workshop of the ITER Expert Group on Confinement and Transport

  6. Results on the ITER Technology R and D

    International Nuclear Information System (INIS)

    1999-01-01

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

  7. Pre-conceptual design assessment of DEMO remote maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Loving, A., E-mail: antony.loving@ccfe.ac.uk [EURATOM/Culham Center Fusion Energy, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Crofts, O.; Sykes, N.; Iglesias, D.; Coleman, M.; Thomas, J. [EURATOM/Culham Center Fusion Energy, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Harman, J. [EFDA Close Support Unit Garching, Boltzmannstaße 2, D-85748 Garching Bei München (Germany); Fischer, U. [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Sanz, J. [Instituto de Fusión Nuclear/UPM, Madrid (Spain); Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Mittwollen, M. [Karlsruhe Institute of Technology, Institut für Fördertechnik und Logistiksysteme, Gotthard-Franz-Straße 8, Geb.50.38, 76131 Karlsruhe (Germany)

    2014-10-15

    EDFA, as part of the Power Plant Physics and Technology programme, has been working on the pre-conceptual design of a Demonstration Power Plant (DEMO). As part of this programme, a review of the remote maintenance strategy considered maintenance solutions compatible with expected environmental conditions, whilst showing potential for meeting the plant availability targets. A key finding was that, for practical purposes, the expected radiation levels prohibit the use of complex remote handling operations to replace the first wall. In 2012/2013, these remote maintenance activities were further extended, providing an insight into the requirements, constraints and challenges. In particular, the assessment of blanket and divertor maintenance, in light of the expected radiation conditions and availability, has elaborated the need for a very different approach from that of ITER. This activity has produced some very informative virtual reality simulations of the blanket segments and pipe removal that are exceptionally valuable in communicating the complexity and scale of the required operations. Through these simulations, estimates of the maintenance task durations have been possible demonstrating that a full replacement of the blankets within 6 months could be achieved. The design of the first wall, including the need to use sacrificial limiters must still be investigated. In support of the maintenance operations, a first indication of the requirements of an Active Maintenance Facility (AMF) has been elaborated.

  8. Software protocol design: Communication and control in a multi-task robot machine for ITER vacuum vessel assembly and maintenance

    International Nuclear Information System (INIS)

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

    2015-01-01

    applying these protocols, the software for a multi-task robot machine that is used for ITER vacuum vessel assembly and maintenance has been developed and it is demonstrated that machining tasks of the robot machine, such as milling, drilling, welding etc., can be implemented in both an individual and composite way.

  9. Software protocol design: Communication and control in a multi-task robot machine for ITER vacuum vessel assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    applying these protocols, the software for a multi-task robot machine that is used for ITER vacuum vessel assembly and maintenance has been developed and it is demonstrated that machining tasks of the robot machine, such as milling, drilling, welding etc., can be implemented in both an individual and composite way.

  10. Occupational Radiation Exposure Analysis of US ITER DCLL TBM

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J; Cadwallader, Lee C; Dagher, Mohamad

    2007-08-01

    This report documents an Occupational Radiation Exposure (ORE) analysis that was performed for the US International Thermonuclear Experimental Reactor (ITER) Dual Coolant Lead Lithium (DCLL) Test Blanket Module (TBM). This analysis was performed with the QADMOD dose code for anticipated maintenance activities for this TBM concept and its ancillary systems. The QADMOD code was used to model the PbLi cooling loop of this TBM concept by specifying gamma ray source terms that simulated radioactive material within the piping, valves, heat exchanger, permeator, pump, drain tank, and cold trap of this cooling system. Estimates of the maintenance tasks that will have to be performed and the time required to perform these tasks where developed based on either expert opinion or on industrial maintenance experience for similar technologies. This report details the modeling activity and the calculated doses for the maintenance activities envisioned for the US DCLL TBM.

  11. Summary report for ITER Task -- D4: Activation calculations for the stainless steel ITER design

    International Nuclear Information System (INIS)

    Attaya, H.

    1995-02-01

    Detailed activation analysis for ITER has been performed as a part of ITER Task D4. The calculations have been performed for the shielding blanket (SS/water) and for the breeding blanket (LiN) options. The activation code RACC-P, which has been modified under IFER Task-D-10 for pulsed operation, has been used in this analysis. The spatial distributions of the radioactive inventory, decay heat, biological hazard potential, and the contact dose were calculated for the two designs for different operation modes and targeted fluences. A one-dimensional toroidal geometrical model has been utilized to determine the neutron fluxes in the two designs. The results are normalized for an inboard and outboard neutron wall loadings of 0.91 and 1.2 MW/M 2 , respectively. The point-wise distributions of the decay gamma sources have been calculated everywhere in the reactor at several times after the shutdown of the two designs and are then used in the transport code ONEDANT to calculate the biological dose everywhere in the reactor. The point-wise distributions of all the responses have also been calculated. These calculations have been performed for neutron fluences of 3.0 MWa/M 2 , which corresponds to the target fluence of ITER, and 0.1 MWa/M 2 , which is anticipated to correspond to the beginning of an extended maintenance period

  12. 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.

  13. 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)

  14. Welding and cutting characteristics of blanket/first wall module to back plate for fusion experimental reactor

    International Nuclear Information System (INIS)

    Sato, Shinichi; Osaki, Toshio; Koga, Shinji

    1996-01-01

    The first wall and the blanket of the International Thermonuclear Experimental Reactor (ITER) are used under severe conditions such as the neutron irradiation by plasma, surface thermal load, the electromagnetic force at the time of plasma disruption and others. Consequently, from the viewpoint of the necessity for disassembling and maintenance, those are divided into modules in toroidal and poloidal directions. In this study, as to the welding of the back plate and the legs supporting blanket modules, which are installed in a vacuum vessel, the characteristic test paying attention to the deformation at the time of welding was carried out, and the optimal welding conditions and the characteristics of welding deformation and others were clarified. Moreover, when water jet method was used for cutting the welded parts of the supporting legs, the properties of the cut parts, the time for cutting and others were examined. The performance required for the welded parts of blanket modules with back plate is shown. The basic test of welding conditions using plate models, partial model test and whole model test are reported. The test of water jet cutting for the maintenance of shielding blanket modules is described. (K.I.)

  15. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    Rocaboy, Alain.

    1982-06-01

    Materials with structures suitable for this purpose are studied. A bibliographic review of the main solid and liquid lithiated compounds is then presented. Erosion, dimensioning and maintenance problems associated with the limiter and the first wall of the reactor are studied from the point of view of the constraints they impose on the design of the blankets. Detailed studies of the main solid and liquid blanket concepts enable the best technological compromises to be determined for the indispensable functions of the blanket to be assured under acceptable conditions. Our analysis leads to four classes of solution, which cannot at this stage be considered as final recommendations, but which indicate what sort of solutions it is worthwhile exploring and comparing in order to be in a position to suggest a realistic blanket at the time when plasma control is sufficiently good for power reactors to be envisaged. Some considerations on the general architecture of the reactor are indicated. Energy storage with pulsed reactors is discussed in the appendix, and a first approach made to minimizing the total tritium recovery [fr

  16. Breeding blanket for Demo

    International Nuclear Information System (INIS)

    Proust, E.; Giancarli, L.

    1992-01-01

    This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently investigated within the framework of the European Test-Blanket Development Programme

  17. ITER ITA newsletter. No. 4, May 2003

    International Nuclear Information System (INIS)

    2003-07-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related meetings, one of them the eighth meeting of the ITER negotiators' standing sub-group (NSSG-8) and a number of related meetings from 14 to 22 May 2003 at Garching, Germany, another was bilateral blanket meeting between ITER International Team (IT) and the Research and Development Institute of Power Engineering (ENTEK), which was held in Moscow, Russian Federation on 22 and 23 May, 2003

  18. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1995-09-01

    There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as Primary Blanket Materials, which have the greatest influence in determining the overall design and performance, and Secondary Blanket Materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified

  19. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1996-01-01

    There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as primary blanket materials, which have the greatest influence in determining the overall design and performance, and secondary blanket materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified. (orig.)

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

    Science.gov (United States)

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

    1993-07-01

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

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

    International Nuclear Information System (INIS)

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

    1993-07-01

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

  2. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

    2016-01-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  3. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

    2016-11-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  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. ITER fuel cycle

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  7. Progress and challenges of the ITER TBM Program from the IO perspective

    International Nuclear Information System (INIS)

    Giancarli, L.M.; Barabash, V.; Campbell, D.J.; Chiocchio, S.; Cordier, J.-J.; Dammann, A.; Dell’Orco, G.; Elbez-Uzan, J.; Fourneron, J.M.; Friconneau, J.P.; Gasparotto, M.; Iseli, M.; Jung, C.-Y.; Kim, B.-Y.; Lazarov, D.; Levesy, B.; Loughlin, M.; Merola, M.; Nevière, J.-C.; Pascal, R.

    2016-01-01

    The paper describes the organization of the Test Blanket Module (TBM) program, its overall objective and schedule and the status of the technical activities within the ITER Organization-Central Team (IO-CT). The latter include the design integration of the Test Blanket Systems (TBSs) into the nuclear buildings, ensuring all interfaces with other ITER systems, the design of the common TBS components such as the TBM Frames, the Dummy TBMs, and the TBS maintenance tools and equipment in the TBM Port Cell as well as in the Hot Cell building, the design of the TBS connection pipes and the definition of the required maintenance operations and associated R&D. The paper also discusses the major challenges that the TBM Program will be facing in ITER such as the potential impact of the TBMs ferritic/martensitic structures on plasma operations, the approaches to tritium and contamination confinement, the required mitigation and recovery actions in case of accidents, and the assessment of the reliability aspects that could have an impact on ITER availability.

  8. Progress and challenges of the ITER TBM Program from the IO perspective

    Energy Technology Data Exchange (ETDEWEB)

    Giancarli, L.M., E-mail: luciano.giancarli@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance (France); Barabash, V.; Campbell, D.J.; Chiocchio, S.; Cordier, J.-J.; Dammann, A.; Dell’Orco, G.; Elbez-Uzan, J.; Fourneron, J.M.; Friconneau, J.P. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance (France); Gasparotto, M. [Max-Planck-Institut für Plasmaphysik, Wendelsteinstraße 1, 17491 Greifswald (Germany); Iseli, M.; Jung, C.-Y.; Kim, B.-Y.; Lazarov, D.; Levesy, B.; Loughlin, M.; Merola, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Pascal, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance (France); and others

    2016-11-01

    The paper describes the organization of the Test Blanket Module (TBM) program, its overall objective and schedule and the status of the technical activities within the ITER Organization-Central Team (IO-CT). The latter include the design integration of the Test Blanket Systems (TBSs) into the nuclear buildings, ensuring all interfaces with other ITER systems, the design of the common TBS components such as the TBM Frames, the Dummy TBMs, and the TBS maintenance tools and equipment in the TBM Port Cell as well as in the Hot Cell building, the design of the TBS connection pipes and the definition of the required maintenance operations and associated R&D. The paper also discusses the major challenges that the TBM Program will be facing in ITER such as the potential impact of the TBMs ferritic/martensitic structures on plasma operations, the approaches to tritium and contamination confinement, the required mitigation and recovery actions in case of accidents, and the assessment of the reliability aspects that could have an impact on ITER availability.

  9. Flow balancing in liquid metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Morley, N.B.

    1995-01-01

    Non-uniform flow distribution between parallel channels is one of the most serious concerns for self-cooled liquid metal blankets with electrically insulated walls. We show that uncertainties in flow distribution can be dramatically reduced by relatively simple design modifications. Several design features which impose flow uniformity by electrically coupling parallel channels are surveyed. Basic mechanisms for ''flow balancing'' are described, and a particular self-regulating concept using discrete passive electrodes is proposed for the US ITER advanced blanket concept. Scoping calculations suggest that this simple technique can be very powerful in equalizing the flow, even with massive insulator failures in individual channels. More detailed analyses and experimental verification will be required to demonstrate this concept for ITER. (orig.)

  10. Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

    International Nuclear Information System (INIS)

    1978-01-01

    This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified

  11. Iter in vessel viewing system design and assessment activities

    Energy Technology Data Exchange (ETDEWEB)

    Neri, C., E-mail: carlo.neri@enea.it [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Costa, P.; Ferri De Collibus, M.; Florean, M.; Mugnaini, G.; Pillon, M.; Pollastrone, F.; Rossi, P. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

    2011-10-15

    The In Vessel Viewing System (IVVS) is 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 prototype of a laser In Vessel Viewing and ranging System was developed and tested at ENEA laboratories in Frascati under EFDA task agreements, it is able to perform sub-millimetric bi-dimensional and three-dimensional images inside ITER during maintenance procedure allowing the evaluation of the state and damages of the in-vessel surface. The present prototype has been designed to operate under room conditions and starting from springtime 2009 a Grant with F4E is in progress for the design and the assessment of the IVVS system for ITER, keeping in account all the environmental conditions and constraints.

  12. 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)

  13. 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)

  14. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Naruse, Yuji; Yamaoka, Mitsuaki; Ohara, Atsushi; Ono, Kiyoshi; Kobayashi, Shigetada.

    1992-06-01

    Aqueous solution blanket using lithium salts such as LiNO 3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

  15. Solid breeder test blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ying, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)]. E-mail: ying@fusion.ucla.edu; Abdou, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Calderoni, P. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Sharafat, S. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Youssef, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); An, Z. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Abou-Sena, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Kim, E. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Reyes, S. [LANL, Livermore, CA (United States); Willms, S. [LANL, Los Alamos, NM (United States); Kurtz, R. [PNNL, Richland, WA (United States)

    2006-02-15

    This paper presents the design and analysis for the US ITER solid breeder blanket test articles. Objectives of solid breeder blanket testing during the first phase of the ITER operation focus on exploration of fusion break-in phenomena and configuration scoping. Specific emphasis is placed on first wall structural response, evaluation of neutronic parameters, assessment of thermomechanical behavior and characterization of tritium release. The tests will be conducted with three unit cell arrays/sub-modules. The development approach includes: (1) design the unit cell/sub-module for low temperature operations and (2) refer to a reactor blanket design and use engineering scaling to reproduce key parameters under ITER wall loading conditions, so that phenomena under investigation can be measured at a reactor-like level.

  16. Fusion fuel blanket technology

    International Nuclear Information System (INIS)

    Hastings, I.J.; Gierszewski, P.

    1987-05-01

    The fusion blanket surrounds the burning hydrogen core of a fusion reactor. It is in this blanket that most of the energy released by the nuclear fusion of deuterium-tritium is converted into useful product, and where tritium fuel is produced to enable further operation of the reactor. As fusion research turns from present short-pulse physics experiments to long-burn engineering tests in the 1990's, energy removal and tritium production capabilities become important. This technology will involve new materials, conditions and processes with applications both to fusion and beyond. In this paper, we introduce features of proposed blanket designs and update and status of international research. In focusing on the Canadian blanket technology program, we discuss the aqueous lithium salt blanket concept, and the in-reactor tritium recovery test program

  17. Blanket testing in NET

    International Nuclear Information System (INIS)

    Chazalon, M.; Daenner, W.; Libin, B.

    1989-01-01

    The testing stages in NET for the performance assessment of the various breeding blanket concepts developed at the present time in Europe for DEMO (LiPb and ceramic blankets) and the requirements upon NET to perform these tests are reviewed. Typical locations available in NET for blanket testing are the central outboard segments and the horizontal ports of in-vessel sectors. These test positions will be connectable with external test loops. The number of test loops (helium, water, liquid metal) will be such that each major class of blankets can be tested in NET. The test positions, the boundary conditions and the external test loops are identified and the requirements for test blankets are summarized (author). 6

  18. A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; Trow, Martin; Dillistone, Michael

    2016-01-01

    'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

  19. Application of remote handling compatibility on ITER plant

    International Nuclear Information System (INIS)

    Sanders, S.; Rolfe, A.; Mills, S.F.; Tesini, A.

    2011-01-01

    The ITER plant will require fully remote maintenance during its operational life. For this to be effective, safe and efficient the plant will have to be developed in accordance with remote handling (RH) compatibility requirements. A system for ensuring RH compatibility on plant designed for Tokamaks was successfully developed and applied, inter alia, by the authors when working at the JET project. The experience gained in assuring RH compatibility of plant at JET is now being applied to RH relevant ITER plant. The methodologies required to ensure RH compatibility of plant include the standardization of common plant items, standardization of RH features, availability of common guidance on RH best practice and a protocol for design and interface review and approval. The protocol in use at ITER is covered by the ITER Remote Maintenance Management System (IRMMS) defines the processes and utilization of management controls including Plant Definition Forms (PDF), Task Definition Forms (TDFs) and RH Compatibility Assessment Forms (RHCA) and the ITER RH Code of Practice. This paper will describe specific examples where the authors have applied the methodology proven at JET to ensure remote handling compatibility on ITER plant. Examples studied are: ·ELM coils (to be installed in-vessel behind the Blanket Modules) - handling both in-vessel, in Casks and at the Hot Cell as well as fully remote installation and connection (mechanical and electrical) in-vessel. ·Neutral beam systems (in-vessel and in the NB Cell) - beam sources, cesium oven, beam line components (accessed in the NB Cell) and Duct Liner (remotely replaced from in-vessel). ·Divertor (in-vessel) - cooling pipe work and remotely operated electrical connector. The RH compatibility process can significantly affect plant design. This paper should therefore be of interest to all parties who develop ITER plant designs.

  20. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

    The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry

  1. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2005-03-01

    Japan Atomic Energy Research Institute (JAERI) has been performing the development of breeding blanket for fusion power plant, as a leading institute of the development of solid breeder blankets, according to the long-term R and D program of the blanket development established by the Fusion Council of Japan in 1999. This report is an overview of development plan, achievements of element technology development and future prospect and plan of the development of the solid breeding blanket in JAERI. In this report, the mission of the blanket development activity in JAERI, key issues and roadmap of the blanket development have been clarified. Then, achievements of the element technology development were summarized and showed that the development has progressed to enter the engineering testing phase. The specific development target and plan were clarified with bright prospect. Realization of the engineering test phase R and D and completion of ITER test blanket module testing program, with universities/NIFS cooperation, are most important steps in the development of breeding blanket of fusion power demonstration plant. (author)

  2. Limitations on blanket performance

    International Nuclear Information System (INIS)

    Malang, S.

    1999-01-01

    The limitations on the performance of breeding blankets in a fusion power plant are evaluated. The breeding blankets will be key components of a plant and their limitations with regard to power density, thermal efficiency and lifetime could determine to a large degree the attractiveness of a power plant. The performance of two rather well known blanket concepts under development in the frame of the European Blanket Programme is assessed and their limitations are compared with more advanced (and more speculative) concepts. An important issue is the question of which material (structure, breeder, multiplier, coatings) will limit the performance and what improvement would be possible with a 'better' structural material. This evaluation is based on the premise that the performance of the power plant will be limited by the blankets (including first wall) and not by other components, e.g. divertors, or the plasma itself. However, the justness of this premise remains to be seen. It is shown that the different blanket concepts cover a large range of allowable power densities and achievable thermal efficiencies, and it is concluded that there is a high incentive to go for better performance in spite of possibly higher blanket cost. However, such high performance blankets are usually based on materials and technologies not yet developed and there is a rather high risk that the development could fail. Therefore, it is explained that a part of the development effort should be devoted to concepts where the materials and technologies are more or less in hand in order to ensure that blankets for a DEMO reactor can be developed and tested in a given time frame. (orig.)

  3. Engineering improvements for the ITER fusion reactor

    International Nuclear Information System (INIS)

    Moreno villar, A.; Albisu, F.

    1996-01-01

    The engineering and design phase (EDA) of the ITER project is being carried out jointly by the European Union, United States, Japan and the Russian Federation. the EU Home Team has contracted the European consortium, EFET, to undertake studies, design tasks and other services relating to the industry's contribution to the overall design of the project. The industry's contribution materializes through work orders within the framework of the contract. This paper summarises the tasks in which EFET participates, especially the ones on which IBERTEF AIE- a company formed by Empresarios Agrupados and SE NER-, collaborates. Up to June 1995, IBERTEF has carried out or is carrying out engineering works in the following tasks: TaskD/: Conceptual design and calculation of the vacuum chamber. Task D8: Structural design of the shielding blanket support. Task D21: Tokamak building crane. Task D51: Power supply schemes. Task CTA T4: Divertor region test platform: Task S22 TD 05: Heavy lifting transporter. Task D71 A-04: Safety design guidelines support. Task G16 TD 17: Design and development of the shielding blanket. Task S74 TD 05: Remote maintenance manual. Task D230: Design of buildings. Task D231: Cost estimate and construction plan for buildings. (Author)

  4. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  5. Blankets for thermonuclear device

    International Nuclear Information System (INIS)

    Maki, Koichi; Fukumoto, Hideshi.

    1986-01-01

    Purpose: To produce tritium more than consumed, through thermonuclear reaction. Constitution: The energy spectrum of neutron generated by neutron multiplying reaction in a neutron multiplying blanket and moderated neutrons has a large ratio in a low energy section. In the low-energy absorption region of stainless steel which is a material of cooling pipes constituting a neutron multiplying blanket cooling channel, the neutrons are absorbed, lessening the neutron multiplying effect. To prevent this, the neutron multiplying blanket cooling channel is covered with tritium breeding blankets, thereby enabling the production of a substantially great amount of tritium more than the amount of tritium to be consumed by the thermonuclear reaction by preventing neutron absorption by the component materials of the cooling channel, improving the tritium breeding ratio by 20 to 25 %, and increasing the efficiency of use of neutrons for tritium generation. (Horiuchi, T.)

  6. Design and development of ceramic breeder demo blanket

    International Nuclear Information System (INIS)

    Enoeda, M.; Sato, S.; Hatano, T.

    2001-01-01

    Ceramic breeder blanket development has been widely conducted in Japan from fundamental researches to project-oriented engineering scaled development. A long term R and D program has been launched in JAERI since 1996 as a course of DEMO blanket development. The objectives of this program are to provide engineering data base and fabrication technologies of the DEMO blanket, aiming at module testing in ITER currently scheduled to start from the beginning of the ITER operation as a near-term target. Two types of DEMO blanket systems, water cooled blanket and helium cooled blanket, have been designed to be consistent with the SSTR (Steady State Tokamak Reactor) which is the reference DEMO reactor design in JAERI. Both of them utilize packed small pebbles of breeder Li 2 O or Li 2 TiO 3 as a candidate) and neutron multiplier (Be) and rely on the development of advanced structural materials (a reduced activation ferritic steel F82H) compatible with high temperature operation. (author)

  7. 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.)

  8. 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.)

  9. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

    Le Marois, G.; Federzoni, L.; Bucci, P.; Revirand, P.

    2000-01-01

    The benefit of HIP techniques applied to the fabrication of fusion internal components for higher performances, reliability and cost savings are emphasized. To demonstrate the potential of the techniques, design of new blankets concepts and mock-ups fabrication are currently performed by CEA. A coiled tube concept that allows cooling arrangement flexibility, strong reduction of the machining and number of welds is proposed for ITER IAM. Medium size mock-ups according to the WCLL breeding blanket concept have been manufactured. The fabrication of a large size mock-up is under progress. These activities are supported by numerical calculations to predict the deformations of the parts during HIP'ing. Finally, several HIP techniques issues have been identified and are discussed

  10. 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.)

  11. Qualification method for a 1 MGy-tolerant front-end chip designed in 65 nm CMOS for the read-out of remotely operated sensors and actuators during maintenance in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Verbeeck, Jens, E-mail: jens.verbeeck@esat.kuleuven.be [KU Leuven (KUL), Div. LRD-MAGyICS, Kasteelpark Arenberg 10, 3001 Heverlee (Belgium); Cao, Ying [KU Leuven (KUL), Div. LRD-MAGyICS, Kasteelpark Arenberg 10, 3001 Heverlee (Belgium); Van Uffelen, Marco; Casellas, Laura Mont; Damiani, Carlo; Morales, Emilio Ruiz; Santana, Roberto Ranz [Fusion for Energy (F4E), c/Josep, no. 2, Torres Diagonal Litoral, Ed. B3, 08019 Barcelona (Spain); Meek, Richard; Haist, Bernhard [Oxford Technologies Ltd. (OTL), 7 Nuffield Way, Abingdon OX14 1RL (United Kingdom); Hamilton, David [ITER Organisation (IO), Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul les Durance Cedex (France); Steyaert, Michiel [KU Leuven, ESAT-MICAS, Kasteelpark Arenberg 10, 3001 Heverlee (Belgium); Leroux, Paul [KU Leuven, ESAT-MICAS, Kasteelpark Arenberg 10, 3001 Heverlee (Belgium); KU Leuven, ESAT, Advanced Integrated Sensing Lab (AdvISe), Kleinhoefstraat 4, 2440 Geel (Belgium)

    2015-10-15

    This paper describes the radiation qualification procedure for a 1 MGy-tolerant Application Specific Integrated Circuit (ASIC) developed in 65 nm CMOS technology. The chip is intended for the read-out of electrical signals of sensors and actuators during maintenance in ITER. First the general working principle of the ASIC is shown. The developed IC allows to read-out, condition and digitize multiple low bandwidth (<10 kHz) sensors. In addition the IC is able to multiplex the digitized sensor signals. To comply with ITER-relevant constraints an adapted radiation qualification procedure has been proposed. The radiation-qualification procedure describes the test criteria and test conditions of the developed ASICs, which are also compared with COTS alternatives, to meet the stringent qualification procedures for electronics exposed to radiation in ITER.

  12. The State of the Art Report on the Development and Manufacturing Technology of Test Blanket Module

    International Nuclear Information System (INIS)

    Lee, J. S.; Jeong, Y. H.; Park, S. Y.; Lee, M. H.; Choi, B. K.; Baek, J. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.; Kim, K. H.

    2006-07-01

    The main objective of the present R and D on breeder blanket is the development of test blanket modules (TBMs) to be installed and tested in International Thermonuclear Experimental Reactor (ITER). In the program of the blanket development, a blanket module test in the ITER is scheduled from the beginning of the ITER operation, and the performance test of TBM in ITER is the most important milestone for the development of the DEMO blanket. The fabrication of TBMs has been required to test the basic performance of the DEMO blanket, i.e., tritium production/recovery, high-grade heat generation and radiation shielding. Therefore, the integration of the TBM systems into ITER has been investigated with the aim to check the safety, reliability and compatibility under nuclear fusion state. For this reason, in the Test Blanket Working Group (TBWG) as an activity of the International Energy Association (IEA), a variety of ITER TBMs have been proposed and investigated by each party: helium-cooled ceramic (WSG-1), helium-cooled LiPb (WSG-2), water-cooled ceramic (WSG-3), self-cooled lithium (WSG-4) and self-cooled molten salt (WSG-5) blanket systems. Because we are still deficient in investigation of TBM development, the need of development became pressing. In this report, for the development of TBM sub-module and mock-up, it is necessary to analyze and examine the state of the art on the development of manufacturing technology of TBM in other countries. And we will be applied as basic data to establish a manufacturing technology

  13. Research and development needs for ITER engineering design

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  16. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  17. Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

    Science.gov (United States)

    Deepak, SHARMA; Paritosh, CHAUDHURI

    2018-04-01

    The Indian test blanket module (TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R&D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices (ITER relevant and DEMO). The Indian Lead–Lithium Cooled Ceramic Breeder (LLCB) blanket concept is one of the Indian DEMO relevant TBM, to be tested in ITER as a part of the TBM program. Helium-Cooled Ceramic Breeder (HCCB) is an alternative blanket concept that consists of lithium titanate (Li2TiO3) as ceramic breeder (CB) material in the form of packed pebble beds and beryllium as the neutron multiplier. Specifically, attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions. These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.

  18. Strategy for the development of EU Test Blanket Systems instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Calderoni, P., E-mail: Pattrick.Calderoni@f4e.europa.eu; Ricapito, I.; Poitevin, Y.

    2013-10-15

    Highlights: • We developed a strategy for the development of instrumentation for EU ITER TBSs. • TBSs instrumentation functions: safety, operation and scientific mission. • Described activities are in support of ITER design review process. -- Abstract: The instrumentation of the HCLL and HCPB Test Blanket System is fundamental in ensuring that ITER safety and operational requirements are satisfied as well as in enabling the scientific mission of the TBM program. It carries out three essential functions: (i) safety, intended as compliance with ITER requirements toward public and workers protection; (ii) system control, intended as compliance with ITER operational requirements and investment protection; and (iii) scientific mission, intended as validating technology and predictive tools for blanket concepts relevant to fusion energy systems. This paper describes the strategy for instrumentation development by providing details of the following five steps to be implemented in procured activities in the short to mid-term (3–4 years): (i) provide mapping of sensors requirements based on critical review of preliminary design data; (ii) develop functional specifications for TBS sensors based on the analysis of operative conditions in the various ITER buildings in which they are located; (iii) assess availability of commercial sensors against developed specifications; (iv) develop prototypes when no available solution is identified; and (v) perform single effect tests for the most critical solicitations and post-test examination of commercial products and prototypes. Examples of technology assessment in two technical areas are included to reinforce and complement the strategy description.

  19. Li2O-pebble type tritium breeding blanket for fusion experimental reactor, 1

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

    1984-01-01

    The fusion experimental reactor is the next stage device in Japan, which is planned to be constructed following the critical plasma experimental device JT-60 being constructed at present. The breeding blanket installed in nuclear fusion reactors is one of most important structures, and it is required to satisfy the fundamental performance of producing and continuously recovering tritium as the nuclear fusion fuel, and other requirement in good coordination. The Li 2 O pebble type breeding blanket that Kawasaki Heavy Industries Ltd. has examined is the concept for resolving the problems of the mass transfer and thermal stress cracking of Li 2 O, which are important in blanket design. In this paper, the concept and characteristics of this breeding blanket are discussed from the viewpoint of the breeding and continuous recovery of tritium, the ease of manufacture and the maintenance of soundness. The breeding blanket is composed of breeding region, tritium purge region, cooling region, plasma stabilizing conductors and blanket container. Li 2 O is excellent in its tritium breeding performance and heat conductivity. The functions required for the breeding blanket, the fundamental structure, the examples of breeding blanket concept, the selection of breeding blanket concept, the characteristics of Li 2 O pebble type blanket and its future prospect are described. (Kako, I.)

  20. ITER EDA status

    International Nuclear Information System (INIS)

    Aymar, R.

    2001-01-01

    '', each representing a potential real procurement contract for an ITER component. The results, after analysis and evaluation by the JCT, have provided the basis for a JCT ''evaluated cost estimates'' report for all packages (Business Confidential) which was presented during a one week meeting at Garching (29 Jan - 2 Feb 2001) to an Ad Hoc Group of Parties' costing experts. The summary was included in the synoptic paper of the PDD for the Council's information. A meeting of the ITER Test Blanket Working Group (TBWG) was held in October 2000. The group has continued its activities during the period of extension of the EDA with a revised charter on the co-ordination of the development work performed by the Parties and by the JCT leading to a co-ordinated test programme on ITER for a DEMO-relevant tritium breeding blanket. This follows earlier work carried out during the EDA, which formed part of the 1998 Final Design Report. For a concise summary of the meeting see the separate article on the Test Blanket Working Group's Recent Activities in the ITER EDA Newsletter, Vol. 10, No. 2, Feb. 2001

  1. ITER EDA Newsletter. V. 6, no. 5

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-05-01

    This issue of the newsletter on Engineering Design Activities (EDA) for the ITER Tokamak project contains a report on Second International Industries` Liaison meeting which was held in Tokyo on 2-4 April 1997 (by Y. Kaneki, JAIF, Japan); an overview report on the Blanket project (by A. Cardella, I.Ioki (ITER Central Team), W. Daenner (EU Home Team)); and a progress report on microwave reflectometry (by J. Sanchez, Madrid, Spain).

  2. ITER EDA Newsletter. V. 6, no. 5

    International Nuclear Information System (INIS)

    1997-05-01

    This issue of the newsletter on Engineering Design Activities (EDA) for the ITER Tokamak project contains a report on Second International Industries' Liaison meeting which was held in Tokyo on 2-4 April 1997 (by Y. Kaneki, JAIF, Japan); an overview report on the Blanket project (by A. Cardella, I.Ioki (ITER Central Team), W. Daenner (EU Home Team)); and a progress report on microwave reflectometry (by J. Sanchez, Madrid, Spain)

  3. Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

    2015-10-15

    Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

  4. Preliminary assessment for dust contamination of ITER In-Vessel Transporter

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    Highlights: •To assess the exposure to the maintenance workers, we calculated the effective dose rate. •To reduce the effective dose rate, the IVT was decontaminated and underwent a design change. •The effective dose rate at each maintenance point was also calculated. -- Abstract: After plasma operations of ITER, radioactive dust will have accumulated in the vacuum vessel (VV). The In-Vessel Transporter (IVT) will be introduced into the VV to remove the shield blanket modules for maintenance or replacement and later reinstall them. The IVT itself also needs to undergo regular maintenance in the Hot Cell Facility (HCF). It is assumed that maintenance workers will be exposed to radioactive dust that has adhered to the surfaces of the IVT. In this study, the areas of the IVT that may be contaminated by dust are evaluated to assess the level of exposure to workers during maintenance work in the HCF. Decontamination processes for the IVT, such as a combination of vacuuming and brushing, were investigated and the dose rate after these processes was evaluated. Even though dust was removed from surfaces where decontamination was possible, the dose rate was very high at some assessment points. To decrease the dose rate in accordance with ALARA policy, a decontamination plan and a maintenance plan, which includes the removal of dust, a radiation shield system, and a reduction in working time are proposed.

  5. 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)

  6. ITER EDA Newsletter. V. 6, no. 8

    International Nuclear Information System (INIS)

    1997-08-01

    This issue of the Newsletter contains a report on blanket remote handling development by K. Shibanuma, R. Haange ( both ITER Joint Central Team) and E. Tada (Japan Home Team) and a report on Fusion research support by the IAEA by T.J. Dolan and U. Schneider Physics Section, IAEA

  7. Investigation on welding and cutting methods for blanket support legs of fusion experimental reactors

    International Nuclear Information System (INIS)

    Tokami, Ikuhide; Nakahira, Masataka; Kurasawa, Toshimasa; Sato, Satoshi; Furuya, Kazuyuki; Hatano, Toshihisa; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1996-07-01

    A toroidally-and poloidally-divided modular blanket has been proposed for a fusion experimental reactor, such as ITER, to enhance its maintainability as well as improve its fabricability. The blanket module, typically the size of 1 m wide, 1-2 m high and 0.4 m deep and the weight of 4 ton, will be supported by support legs which are extruded from back of the module and connected to a 70-100 mm thick strong back plate. The support leg has to withstand large electromagnetic force during plasma disruption and provide the way for in-situ module replacement by remote handling. For the connection method of the support leg to the back plate, a welding approach has been investigated here in terms of its high reliability against the large electromagnetic loads. For the welding approach, the support leg needs to be 70 mm thick, and the working space for welding/cutting heads are limited to 100 mm x 150 mm adjacent to the support leg. Based on a comparison of several welding methods, e.g. NGTIG, NGMIG and laser, NGTIG has been selected as a reference due to its well-established technology and the least R and D required. As for the cutting method, a plasma cutting has been given the highest priority to be pursued because of its compactness and high speed. Through preliminary design studies, the possibility of small welding/cutting heads that will work in the limited space has been shown, and maintenance route for in-situ module replacement with pre-and postfixture of the module has been investigated. Also preliminary R and Ds have resulted in; 1)the welding distortion is predictable according to the shape of weld groove and adjustable to meet the placement requirement of the module first wall, 2)the plasma cut surface can be rewelded without machining, 3)the welding/cutting time will meet the requirement of maintenance time. (author)

  8. Maintenance of a commercial fusion power station and its implications for safety

    International Nuclear Information System (INIS)

    Sherwood, D.V.; Pearcey, J.; Thompson, H.M.; Sublet, J.-Ch.; Taylor, N.P.

    1994-11-01

    A recent study was undertaken to identify a maintenance scheme which would lead to an acceptable level of availability for a future commercial fusion power station. It was thought that a scheme involving the servicing of the divertor through dedicated maintenance ports provided in each sector of the torus and the sub-division of the blanket in both the radial and poloidal directions to allow it to be handled through two radial access ports, could lead to a plant availability in the order of 80% (including 7% unplanned shutdown) i.e. comparable with that expected from APWRs. The method of blanket replacement proposed is radically different from that of the ITER CDA. The maintenance scheme has now been developed sufficiently to allow the main potential hazards requiring control during maintenance to be identified. Taking these into account, a more accurate assessment of the plant availability has been possible which confirms that 80% should be possible. The scheme is considered to present little risk to the maintenance staff and with proper safeguards there would be little potential for the spread of contamination. (author)

  9. Integration of test modules in the main blanket and vacuum vessel design

    International Nuclear Information System (INIS)

    Nakahira, Masataka; Kurasawa, Toshimasa; Sato, Satoshi; Furuya, Kazuyuki; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1995-07-01

    Typical test modules for water-cooled and helium-cooled ceramic breeder blankets have been designed, and their major design parameters are summarized. Among various candidates studied in Japan at present, BOT (Breeder Out of Tube) type of blanket is exemplified here. The integration scheme of the test module into ITER basic machine is also shown. Even with other type of blanket, the integration scheme won't be affected. The composition and space requirement of cooling and tritium recovery systems for the test module have also been studied. (author)

  10. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    Ozawa, Yoshihiro; Uda, Tatsuhiko; Maki, Koichi.

    1993-01-01

    The present invention provides a blanket of a thermonuclear device which produces tritium fuels consumed in plasmas while converting neutrons generated in the plasmas into heat energy. That is, zirconium is coated to at least one of neutron breeder pebbles and breeder pebbles, to suppress reaction between them by being in direct contact with each other at a high temperature. Further, fins are attached to a cooling pipe at a pitch smaller than the diameter of both of the pebbles, to prevent direct contact at whole surface of the pebbles and the cooling pipe, which would lower a temperature excessively. The length of the fin is controlled to control the thickness of a helium gas gap. With such constitution, direct contact of neutron breeder pebbles and the breeder pebble which are to be filled and mixed, and tend to react at a high temperature, can be prevented. The temperature of the breeding blanket is reliably prevented from lowering below a tritium emitting temperature. The structure is simplified and the production is facilitated. (I.S.)

  11. 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

  12. Neutron cameras for ITER

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  13. Design study of blanket structure based on a water-cooled solid breeder for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Youji; Tobita, Kenji; Utoh, Hiroyasu; Tokunaga, Shinji; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Sakamoto, Yoshiteru

    2015-10-15

    Highlights: • Neutronics design of a water-cooled solid mixed breeder blanket was presented. • The blanket concept achieves a self-sufficient supply of tritium by neutronics analysis. • The overall outlet coolant temperature was 321 °C, which is in the acceptable range. - Abstract: Blanket concept with a simplified interior for mass production has been developed using a mixed bed of Li{sub 2}TiO{sub 3} and Be{sub 12}Ti pebbles, coolant conditions of 15.5 MPa and 290–325 °C and cooling pipes without any partitions. Considering the continuity with the ITER test blanket module option of Japan and the engineering feasibility in its fabrication, our design study focused on a water-cooled solid breeding blanket using the mixed pebbles bed. Herein, we propose blanket segmentation corresponding to the shape and dimension of the blanket and routing of the coolant flow. Moreover, we estimate the overall tritium breeding ratio (TBR) with a torus configuration, based on the segmentation using three-dimensional (3D) Monte Carlo N-particle calculations. As a result, the overall TBR is 1.15. Our 3D neutronics analysis for TBR ensures that the blanket concept can achieve a self-sufficient supply of tritium.

  14. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    1983-10-01

    This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies

  15. The Test Blanket Modules project in Europe: From the strategy to the technical plan over next ten years

    International Nuclear Information System (INIS)

    Poitevin, Y.; Zmitko, M.; Orco, G. dell; Laesser, R.; Diegele, E.; Sundstroem, J.; Boccaccini, L.; Salavy, J.-F.

    2006-01-01

    The testing of Breeding Blanket concepts in ITER is recognized as an essential milestone in the development of a future reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeding blankets for DEMO reactor specifications that will be tested in ITER: the Helium-Cooled Lithium-Lead (HCLL) blanket which uses the eutectic Pb-15. 7 Li as both breeder and neutron multiplier, and the Helium-Cooled Pebble-Bed (HCPB) blanket which features lithiated ceramic pebbles (Li 4 SiO 4 or Li 2 TiO 3 ) as breeder and beryllium pebbles as neutron multiplier. Both blankets are using the pressurized He technology for heat extraction (8 MPa, inlet/outlet temperature 300/500 o C) and a 9% CrWVTa Reduced Activation Ferritic Martensitic (RAFM) steel as structural material, the EUROFER. Referring to the so called '' fast-track '' EU scenario, those concepts are intended to be tested in ITER, getting the maximum of information required for launching the DEMO blanket design and construction after the first 10 years of ITER operation. For that, the EU has adopted a blanket testing strategy based on the development of Test Blanket Modules (TBMs) that are expected to use DEMO relevant technologies and are designed for each ITER plasma phase to optimize the feedback and to avoid any impact on ITER availability. Following the decision on ITER construction, the EU has reviewed and detailed the fundamental elements for an implementation of the future EU TBMs Project aimed at delivering TBMs Systems to ITER under suitable schedule and acceptance standards. For that the following items have been analyzed in detail and are reported in the present paper: · Impact of the ITER environment (design, standards, schedule, operational scheme) on the TBM systems design and development plan · Project technical plan with focus on the next ten years up to the installation of the first TBMs in ITER · Project risk

  16. 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...

  17. Critical Design Factors for Sector Transport Maintenance in DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Utoh, H.; Someya, Y.; Tobita, K.; Asakura, N.; Hoshino, K.; Nakamura, M., E-mail: uto.hiroyasu@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

    2012-09-15

    Full text: Maintenance is a critical issue for fusion DEMO reactor because the design conditions and requirements of DEMO maintenance scheme are different from that of ITER remote handling. The sector transport maintenance scheme has advantages to maintain blankets and divertors without the use of sophisticated remote handling devices including sensitive devices to radiation in the reactor. SlimCS designed in JAEA adopts the sector transport maintenance scheme in which every sector is pulled out horizontally through a port between TF coils. A critical design issue for the horizontal sector transport maintenance scheme is how to support an enormous turnover force of the toroidal field (TF) coils. We propose following two options; first option is the horizontal transport maintenance scheme in which every sector is pulled out through four horizontal ports connected with the corridor. Second option is the vertical sector transport maintenance scheme with small vertical maintenance ports (total: 6 ports). The new horizontal sector transport limited in the number of maintenance ports is a more realistic maintenance scheme, and the key engineering issue is the transferring mechanism of sector in the vacuum vessel. In the maintenance scenario, the key design factors are the cool down time in reactor and the cooling method in maintenance scheme for keeping components under operation temperature. By one-dimensional heat conduction analysis, the sector should be transported to hot cell within 40 hours in the case the cool down time is one month. In the horizontal sector transport maintenance, the maintenance time including removal of cooling piping, drain of cooling water and sector transport to hot cell is about 32 hours. Furthermore, the tritium release in the sector transport can be suppressed because the components temperature drops by forced-air cooling system. This paper mainly focuses on a sector transport maintenance scheme from the aspects of high plant availability

  18. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  19. Predictive capabilities, analysis and experiments for Fusion Nuclear Technology, and ITER R ampersand D

    International Nuclear Information System (INIS)

    1991-01-01

    This report discusses the following topics on ITER research and development: trituim modeling; liquid metal blanket modeling; free surface liquid metal studies; and thermal conductance and thermal control experiments and modeling

  20. The design of the ITER first wall panels

    Energy Technology Data Exchange (ETDEWEB)

    Mitteau, R., E-mail: raphael.mitteau@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Calcagno, B.; Chappuis, P.; Eaton, R.; Gicquel, S. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Chen, J. [Southwestern Institute of Physics, Huangjing Road, Chengdu 610225 (China); Labusov, A. [Efremov Research Institute, 189631 St. Petersburg (Russian Federation); Martin, A.; Merola, M.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ulrickson, M. [Sandia National Laboratory, Albuquerque, NM (United States); Zacchia, F. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: • The ITER blanket is in the final stage of design completion. • Issues raised about the blanket heat loads and remote handling strategy are addressed, while integrating the in-vessel coils. • Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R and D activities. -- Abstract: The ITER blanket is in the final stage of design completion. The issues raised during the 2007 ITER design review about the first wall (FW) heat loads and remote handling strategy have been addressed, while integrating the recently confirmed in-vessel coils. This paper focuses on the FW design, which is nearing completion. Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R and D activities.

  1. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2010-06-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  2. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2007-08-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  3. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    International Nuclear Information System (INIS)

    Lee C. Cadwallader

    2007-01-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with 'generic' component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance

  4. 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

  5. An Analysis of Ripple and Error Fields Induced by a Blanket in the CFETR

    Science.gov (United States)

    Yu, Guanying; Liu, Xufeng; Liu, Songlin

    2016-10-01

    The Chinese Fusion Engineering Tokamak Reactor (CFETR) is an important intermediate device between ITER and DEMO. The Water Cooled Ceramic Breeder (WCCB) blanket whose structural material is mainly made of Reduced Activation Ferritic/Martensitic (RAFM) steel, is one of the candidate conceptual blanket design. An analysis of ripple and error field induced by RAFM steel in WCCB is evaluated with the method of static magnetic analysis in the ANSYS code. Significant additional magnetic field is produced by blanket and it leads to an increased ripple field. Maximum ripple along the separatrix line reaches 0.53% which is higher than 0.5% of the acceptable design value. Simultaneously, one blanket module is taken out for heating purpose and the resulting error field is calculated to be seriously against the requirement. supported by National Natural Science Foundation of China (No. 11175207) and the National Magnetic Confinement Fusion Program of China (No. 2013GB108004)

  6. Activation and afterheat analyses for the HCPB test blanket

    International Nuclear Information System (INIS)

    Pereslavtsev, P.; Fischer, U.

    2007-01-01

    The Helium-Cooled Pebble Bed (HCPB) blanket is one of two breeder blanket concepts developed in the framework of the European Fusion Technology Programme for performance tests in ITER. The recent development programme focussed on the detailed engineering design of the Test Blanket Module (TBM) and associated systems including the assessment of safety and licensing related issues with the objective to prepare for a preliminary Safety Report. To provide a sound data basis for the safety analyses of the HCPB TBM system in ITER, the afterheat and activity inventories were assessed making use of a code system that allows performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. A suitable MCNP model of a 20 degree ITER torus sector with an integrated TBM of the HCPB PI (Plant Integration) type in the horizontal test blanket port was developed and adapted to the requirements for coupled 3D neutron transport and activation calculations. Two different irradiation scenarios were considered in the coupled 3D neutron transport and activation calculations. The first one is representative for the TBM irradiation in ITER with a total of 9000 neutron pulses over a three (calendar) years period. It was simulated by a continuous irradiation for 3 years minus the last month and a discontinuous irradiation with 250 pulses (420 s pulse length, 1200 s power-off in between) over the last month. The second (conservative) irradiation scenario assumes an extended irradiation time over the full anticipated lifetime of ITER according to the M-DRG-1 irradiation scenario with a total first wall fluence of 0.3 MWa/m 2 . For both irradiation scenarios the radioactivity inventories, the afterheat and the contact gamma dose were calculated as function of the decay time. Data were processed for the total activity and afterheat of the TBM, its constituting components and materials including their

  7. ITER project and fusion technology

    International Nuclear Information System (INIS)

    Takatsu, H.

    2011-01-01

    In the sessions of ITR, FTP and SEE of the 23rd IAEA Fusion Energy Conference, 159 papers were presented in total, highlighted by the remarkable progress of the ITER project: ITER baseline has been established and procurement activities have been started as planned with a target of realizing the first plasma in 2019; ITER physics basis is sound and operation scenarios and operational issues have been extensively studied in close collaboration with the worldwide physics community; the test blanket module programme has been incorporated into the ITER programme and extensive R and D works are ongoing in the member countries with a view to delivering their own modules in a timely manner according to the ITER master schedule. Good progress was also reported in the areas of a variety of complementary activities to DEMO, including Broader Approach activities and long-term technology. This paper summarizes the highlights of the papers presented in the ITR, FTP and SEE sessions with a minimum set of background information.

  8. Remote handling assessment of attachment concepts for DEMO blanket segments

    Energy Technology Data Exchange (ETDEWEB)

    Iglesias, Daniel, E-mail: daniel.iglesias@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Bastow, Roger; Cooper, Dave; Crowe, Robert; Middleton-Gear, Dave [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Sibois, Romain [VTT, Technical Research Centre of Finland, Industrial Systems, ROViR, Tampere (Finland); Carloni, Dario [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT) (Germany); Vizvary, Zsolt; Crofts, Oliver [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Harman, Jon [EFDA Close Support Unit Garching, Boltzmannstaße 2, D-85748 Garching bei München (Germany); Loving, Antony [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2015-10-15

    Highlights: • Challenges are identified for the remote handling of blanket segments’ attachments. • Two attachment design approaches are assessed for remote handling (RH) feasibility. • An alternative is proposed, which potentially simplifies and speeds-up RH operations. • Up to three different assemblies are proposed for the remote handling of the attachments. • Proposed integrated design of upper port is compatible with the attachment systems. - Abstract: The replacement strategy of the massive Multi-Module Blanket Segments (MMS) is a key driver in the design of several DEMO systems. These include the blankets themselves, the vacuum vessel (VV) and its ports and the Remote Maintenance System (RMS). Common challenges to any blanket attachment system have been identified, such as the need for applying a preload to the MMS manifold, the effects of the decay heat and several uncertainties related to permanent deformations when removing the blanket segments after service. The WP12 kinematics of the MMS in-vessel transportation was adapted to the requirements of each of the supports during 2013 and 2014 design activities. The RM equipment envisaged for handling attachments and earth connections may be composed of up to three different assemblies. An In-Vessel Mover at the divertor level handles the lower support and earth bonding, and could stabilize the MMS during transportation. A Shield Plug crane with a 6 DoF manipulator operates the upper attachment and earth straps. And a Vertical Maintenance Crane is responsible for the in-vessel MMS transportation and can handle the removable upper support pins. A final proposal is presented which can potentially reduce the number of required systems, at the same time that speeds-up the RMS global operations.

  9. Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

    2016-11-15

    Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

  10. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    Smith, D.L.; Sze, D.K.

    1986-02-01

    The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

  11. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

    Sze, D.K.; Jung, J.; Cheng, E.T.

    1986-01-01

    Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

  12. The blanket interface to TSTA

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Grimm, T.L.; Sze, D.K.; Anderson, J.L.; Bartlit, J.R.; Naruse, Y.; Yoshida, H.

    1988-01-01

    The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction

  13. Evaluation of induced activity, decay heat and dose rate distribution after shutdown in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Maki, Koichi [Hitachi Ltd., Ibaraki (Japan). Hitachi Research Lab.; Satoh, Satoshi; Hayashi, Katsumi; Yamada, Koubun; Takatsu, Hideyuki; Iida, Hiromasa

    1997-03-01

    Induced activity, decay heat and dose rate distributions after shutdown were estimated for 1MWa/m{sup 2} operation in ITER. The activity in the inboard blanket one day after shutdown is 1.5x10{sup 11}Bq/cm{sup 3}, and the average decay heating rate 0.01w/cm{sup 3}. The dose rate outside the 120cm thick concrete biological shield is two order higher than the design criterion of 5{mu}Sv/h. This indicates that the biological shield thickness should be enhanced by 50cm in concrete, that is, total thickness 170cm for workers to enter the reactor room and to perform maintenance. (author)

  14. 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)

  15. 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)

  16. European TBM for ITER: Structural material assessment and breeding capability - Comparative analysis

    International Nuclear Information System (INIS)

    Herreras, Y.; Perlado, J.M.; Ibarra, A.

    2007-01-01

    Full text of publication follows: The ITER European Party is currently developing for DEMO reactor specifications two breeding blanket concepts: the Helium-Cooled Lithium-Lead blanket (HCLL), using a liquid breeder; and the Helium-Cooled Pebble-Bed blanket (HCPB), using a lithiated solid breeder. These two research lines are expected to be tested in ITER as Test Blanket Modules (TBM), in order to demonstrate their safety, economical and environmental suitability. In this sense, structural material activation and breeding blanket capability represent two major challenges. This paper presents new calculations regarding neutronic irradiation inside the ITER Vacuum Vessel. In particular, results are focused on the irradiation affecting the equatorial ports, where the TBM will be located for testing. The methodology employed mainly consists in calculating the neutronic irradiation levels at the required locations with the transport code MCNP, where the input geometry has been previously designed with the program CATIA V5. The main structural materials proposed for the European Test blanket Modules are selected in order to carry out a comparative analysis in safety terms: material activation and basic parameters for damage analysis are evaluated with the code ACAB, based on the neutronic irradiation results mentioned above. Finally, the breeding blanket capability is assessed for both breeding blanket concepts; the results are compared considering the choice of the structural material. (authors)

  17. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi; Nagakura, Masaaki; Kanzawa, Toru.

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman's equation within +25 ∼ -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author)

  18. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Rampal, G.; Boccaccini, L.V.; Meyder, R.; Neuberger, H.; Laesser, R.; Poitevin, Y.; Zmitko, M.; Rigal, E.

    2006-01-01

    The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R-and-D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

  19. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-06-01

    The fusion blanket surrounds the burning hydrogen core of the fusion reactor. It is in this blanket that most of the energy released by the DT fusion reaction is converted into useable product, and where tritium fuel is produced to enable further operation of the reactor. Blankets will involve new materials, conditions and processes. Several recent fusion blanket concepts are presented to illustrate the range of ideas

  20. Alara applied to iter design and operation

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  1. 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

  2. An Asdex-type divertor for ITER

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1989-01-01

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

  3. The ITER EC H and CD upper launcher: Analysis of remote handling compatibility

    International Nuclear Information System (INIS)

    Ronden, D.M.S.; Baar, M. de; Chavan, R.; Elzendoorn, B.S.Q.; Goodman, T.; Heemskerk, C.J.M.; Henderson, M.A.; Koning, J.F.; Saibene, G.; Spaeh, P.; Strauss, D.

    2011-01-01

    Research highlights: → RH class 1 requires a full RH compatible design and a detailed maintenance plan that needs to be demonstrated through hardware mockup testing. → RH class 2 requires a full RH compatible design and a detailed and verified maintenance plan. → RH class 3 requires a RH compatible design and a basic maintenance plan. - Abstract: The present design of the ECH (Electron Cyclotron Heating) upper port launcher has been evaluated in light of the ITER remote handling (RH) requirements. Changes to the launcher design associated with the accessibility, maintainability and manageability of replaceable components are presented. Captive bolts were placed along the flange of the Blanket Shielding Module (BSM). A hinge mechanism was integrated to simplify the (dis-)mounting of the BSM and a frame with incorporated cooling and actuation lines was suggested for simplified mounting and replacement of the steerable mirrors. Rotating the upper port plug upside-down improves maintenance access and component handling. Tools are proposed for manipulation of the port plug and its sub-components. The RH compatibility analysis can improve a design. Early consideration of RH requirements and implementation of necessary features is therefore vital.

  4. Blanket Manufacturing Technologies : Thermomechanical Tests on HCLL Blanket Mocks Up

    International Nuclear Information System (INIS)

    Laffont, G.; Cachon, L.; Taraud, P.; Challet, F.; Rampal, G.; Salavy, J.F.

    2006-01-01

    In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the '' HCLL blanket concept '' have motivated the present study. The aim of this study, carried out in the frame of EFDA Work program, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-up under breeder blanket representative operating conditions.The first step of this experimental program is the design and manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling loop (pressure of 80 bar, maximum temperature of 500 o C,flow rate of 30 g/s) taking the opportunity of synergies with the gas-cooled fission reactor R-and-D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the program made to support the cooling plate mock up tests which will be carried out on the DIADEMO facility (CEA) by thermo-mechanical calculations in order to define the relevant test conditions and the experimental parameters to be monitored. (author)

  5. U.S. Contributions to ITER

    International Nuclear Information System (INIS)

    Sauthoff, Ned R.

    2005-01-01

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

  6. ITER hydrogen isotope separation system conceptual design description

    International Nuclear Information System (INIS)

    Busigin, A.; Sood, S.K.; Kveton, O.K.; Dinner, P.J.; Murdoch, D.K.; Leger, D.

    1990-01-01

    This paper presents integrated hydrogen Isotope Separation System (ISS) designs for ITER based on requirements for plasma exhaust processing, neutral beam injection deuterium cleanup, pellet injector propellant detritiation, waste water detritiation, and breeding blanket detritiation. Specific ISS designs are developed for a machine with an aqueous lithium salt blanket (ALSB) and a machine with a solid ceramic breeding blanket (SBB). The differences in the ISS designs arising from the different blanket concepts are highlighted. It is found that the ISS designs for the two blanket concepts considered are very similar with the only major difference being the requirement for an additional large water distillation column for ALSB water detritiation. The extraction of tritium from the ALSB is based on flash evaporation to separate the blanket water from the dissolved Li salt, with the tritiated water then being fed to the ISS for detritiation. This technology is considered to be relatively well understood in comparison to front-end processes for SBB detritiation. In the design of the cryogenic distillation portion of the ISS, it was found that the tritium inventory could be very large (> 600 g) unless specific design measures were taken to reduce it. In the designs which are presented, the tritium inventory has been reduced to about 180 g, which is less than the ITER single-failure release limit of 200 g. Further design optimization and isolation of components is expected to reduce the inventory further. (orig.)

  7. 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)

  8. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2014-01-01

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM

  9. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tariq Siddique, M.; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-05-15

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM.

  10. Non-linear failure analysis of HCPB blanket for DEMO taking into account high dose irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Aktaa, J., E-mail: jarir.aktaa@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Kecskés, S.; Pereslavtsev, P.; Fischer, U.; Boccaccini, L.V. [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-10-15

    Highlights: • First non-linear structural analysis for the European Helium Cooled Pebble Bed Blanket Module taking into account high dose irradiation. • Most critical areas were identified and analyzed with regard to the effect of irradiation on predicted damage at these areas. • Despite the extensive computing time 100 cycles were simulated by using the sub-modelling technique investigating damage at most critical area. • The results show a positive effect of irradiation on calculated damage which is mainly attributed to the irradiation induced hardening. - Abstract: For the European helium cooled pebble bed (HCPB) blanket of DEMO the reduced activation ferritic martensitic steel EUROFER has been selected as structural material. During operation the HCPB blanket will be subjected to complex thermo-mechanical loadings and high irradiation doses. Taking into account the material and structural behaviour under these conditions is a precondition for a reliable blanket design. For considering high dose irradiation in structural analysis of the DEMO blanket, the coupled deformation damage model, extended recently taking into account the influence of high dose irradiation on the material behaviour of EUROFER and implemented in the finite element code ABAQUS, has been used. Non-linear finite element (FE) simulations of the DEMO HCPB blanket have been performed considering the design of the HCPB Test Blanket Module (TBM) as reference and the thermal and mechanical boundary conditions of previous analyses. The irradiation dose rate required at each position in the structure as an additional loading parameter is estimated by extrapolating the results available for the TBM in ITER scaling the value calculated in neutronics and activation analysis for ITER boundary conditions to the DEMO boundary conditions. The results of the FE simulations are evaluated considering damage at most critical highly loaded areas of the structure and discussed with regard to the impact of

  11. Non-linear failure analysis of HCPB blanket for DEMO taking into account high dose irradiation

    International Nuclear Information System (INIS)

    Aktaa, J.; Kecskés, S.; Pereslavtsev, P.; Fischer, U.; Boccaccini, L.V.

    2014-01-01

    Highlights: • First non-linear structural analysis for the European Helium Cooled Pebble Bed Blanket Module taking into account high dose irradiation. • Most critical areas were identified and analyzed with regard to the effect of irradiation on predicted damage at these areas. • Despite the extensive computing time 100 cycles were simulated by using the sub-modelling technique investigating damage at most critical area. • The results show a positive effect of irradiation on calculated damage which is mainly attributed to the irradiation induced hardening. - Abstract: For the European helium cooled pebble bed (HCPB) blanket of DEMO the reduced activation ferritic martensitic steel EUROFER has been selected as structural material. During operation the HCPB blanket will be subjected to complex thermo-mechanical loadings and high irradiation doses. Taking into account the material and structural behaviour under these conditions is a precondition for a reliable blanket design. For considering high dose irradiation in structural analysis of the DEMO blanket, the coupled deformation damage model, extended recently taking into account the influence of high dose irradiation on the material behaviour of EUROFER and implemented in the finite element code ABAQUS, has been used. Non-linear finite element (FE) simulations of the DEMO HCPB blanket have been performed considering the design of the HCPB Test Blanket Module (TBM) as reference and the thermal and mechanical boundary conditions of previous analyses. The irradiation dose rate required at each position in the structure as an additional loading parameter is estimated by extrapolating the results available for the TBM in ITER scaling the value calculated in neutronics and activation analysis for ITER boundary conditions to the DEMO boundary conditions. The results of the FE simulations are evaluated considering damage at most critical highly loaded areas of the structure and discussed with regard to the impact of

  12. 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

  13. Quantification and disposal of radioactive waste from ITER operation

    International Nuclear Information System (INIS)

    Olsson, G.; Devell, L.; Johnsson, B.; Gulden, W.

    1991-01-01

    The work on the safety and environment for the Next European Torus (NET) is being performed within the European Fusion Technology Safety and Environment Programme by the NET team and under NET contracts. In the area of NET-oriented investigations concerning waste management and disposal, Studsvik is concentrating on the operational waste from both NET and ITER (International Thermonuclear Experimental Reactor). This paper gives a characterization and quantification of the radioactive waste generated from the operation of ITER during the Physics Phase, and from the replacement of all blanket segments (European shielding blanket option) at the end of the Physics Phase after an integrated first-wall loading of 0.03 MWy/m 2 . The total activity contents and volumes of packaged waste from the Physics Phase operation and from the blanket replacement are estimated. The waste volume from replacement of the shielding blanket segments of ITER is considerably larger than estimated in earlier calculations for NET due to the fact that the ITER conceptual design includes more of the stell shielding in the removable segments. The waste handling and disposal are described using existing Swedish and German concepts for similar waste categories from nuclear fission reactors. This includes the choice of suitable packagings, intermediate storage time for cooling, and type of repository for final disposal. Some typical cost figures for waste handling are also presented. (orig.)

  14. Tritium module for ITER/Tiber system code

    International Nuclear Information System (INIS)

    Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

    1988-01-01

    A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs

  15. Transmutation blanket design for a Tokamak system

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A. Fortini; Costa, Antonella L.

    2011-01-01

    Sub-critical advanced reactor with a D-T fusion neutron source based on Tokamak technology is an innovative type of nuclear system. Due to the high quantity of neutrons produced by fusion reactions, it could be well spent in the transmutation process of the transuranic elements. Nevertheless, to achieve a successful transmutation, it is necessary to know the neutron fluence along the radial axis and its characteristics. In this work, it evaluated the neutron flux and interaction frequency along the radial axis changing the material of the first wall. W-alloy, beryllium and the combination of both were studied and regions more suitable to transmutation were determined. The results demonstrated that the better zone to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements W-alloy/W-alloy and W-alloy/Beryllium would be able to hold the requirements of high fluence and hardening spectrum needed to transuranic transmutation. The system was simulated using the MCNP5 code, the ITER Final Design Report, 2001, and the FENDL/MC-2.1 nuclear data library. (author)

  16. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1979-11-01

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  17. ITER cooling system

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-11-01

    The present specification of the ITER cooling system does not permit its operation with water above 150 C. However, the first wall needs to be heated to higher temperatures during conditioning at 250 C and bake-out at 350 C. In order to use the cooling water for these operations the cooling system would have to operate during conditioning at 37 Bar and during bake-out at 164 Bar. This is undesirable from the safety analysis point of view, and alternative heating methods are to be found. This review suggests that superheated steam or gas heating can be used for both baking and conditioning. The blanket design must consider the use of dual heat transfer media, allowing for change from one to another in both directions. Transfer from water to gas or steam is the most intricate and risky part of the entire heating process. Superheated steam conditioning appears unfavorable. The use of inert gas is recommended, although alternative heating fluids such as organic coolant should be investigated

  18. 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

  19. Neutronic design and performance analysis of Korean ITER TBM by Monte Carlo method

    International Nuclear Information System (INIS)

    Kim, Chang Hyo; Han, Beom Seok; Park, Ho Jin

    2006-01-01

    The objective of this project is to develop a neutronic design of the Korean TBM(Test Blanket Module) which will be installed in ITER(International Thermonuclear Experimental Reactor). This project is intended to analyze a neutronic design and nuclear performances of the Korean ITER TBM through the transport calculation of MCCARD. In detail, we will conduct numerical experiments for developing the neutronic design of the Korean ITER TBM and improving the nuclear performances. The results of the numerical experiments produced in this project will be utilized for a design optimization of the Korean ITER TBM. In this project, we proposed the neutronic methodologies for analyzing the nuclear characteristics of the fusion blanket. In order to investigate the behavior of neutrons and photons in the fusion blanket, Monte Carlo transport calculation was conducted with MCCARD. In addition, to optimize the neutronic performances of the fusion blanket, we introduced the design concept using a graphite reflector and a Pb multiplier. Through various numerical experiments, it was verified that these design concepts can be utilized efficiently to improve neutronic performances and resolve many drawbacks. The graphite-reflected HCML blanket can provide the neutronic performances far better than the non-reflected blanket, and a slightly-enriched Li breeder can satisfy the tritium self-sufficiency. The HCSB blanket design concept with a graphite reflector and a Pb multiplier was proposed. According to results of the neutronic analyses, the graphite-reflected HCSB blanket with a Pb multiplier can provide the neutronic performances comparable with those of the conventional HCSB blanket

  20. 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)

  1. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    1983-10-01

    The objectives of the Blanket Comparison and Selection Study (BCSS) can be stated as follows: (1) Define a small number (approx. 3) of blanket design concepts that should be the focus of the blanket R and D program. A design concept is defined by the selection of all materials (e.g., breeder, coolant, structure and multiplier) and other major characteristics that significantly influence the R and D requirements. (2) Identify and prioritize the critical issues for the leading blanket concepts. (3) Provide the technical input necessary to develop a blanket R and D program plan. Guidelines for prioritizing the R and D requirements include: (a) critical feasibility issues for the leading blanket concepts will receive the highest priority, and (b) for equally important feasibility issues, higher R and D priority will be given to those that require minimum cost and short time

  2. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

    2000-07-01

    This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

  3. ITER Fast Ion Collective Thomson Scattering

    DEFF Research Database (Denmark)

    Bindslev, Henrik; Larsen, Axel Wright; Meo, Fernando

    2005-01-01

    The EFDA Contract 04-1213 with Risø National Laboratory concerning a detailed integrated design of a Fast Ion Collective Thomson Scattering (CTS) diagnostic for ITER was signed on 31 December 2004. In 2003 the Risø CTS group finished a feasibility study and a conceptual design of an ITER Fast Ion...... Collective Thomson Scattering System (Contract 01.654) [1, 2]. The purpose of the CTS diagnostic is to measure the distribution function of fast ions in the plasma. The feasibility study demonstrated that the only system that can fully meet the ITER measurement requirements for confined fusion alphas is a 60...... the blanket gap, and calculations of diagnosing fuel ion ratio and rotation velocity by CTS....

  4. 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.

  5. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

    Malang, S.; Casal, V.; Arheidt, K.; Fischer, U.; Link, W.; Rust, K.

    1986-01-01

    A blanket concept for NET using liquid lithium-lead both as breeder material and as coolant is described. The need for inboard breeding is avoided by using beryllium as neutron multiplier in the outboard blanket. Novel flow channel inserts are employed in all poloidal ducts to reduce the MHD pressure drop. The concept offers a simple mechanical design and a higher tritium breeding ratio compared to water- and gas-cooled blankets. (author)

  6. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

    Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Usher, J.L.

    1980-04-01

    Definitions are given of 10 generic blanket types and the specific blanket chosen to be analyzed in detail from each of the 10 types. Dimensions, compositions, energy depositions and breeding ratios (where applicable) are presented for each of the 10 designs. Ultimately, based largely on neutronics and thermal hyraulics results, breeding an nonbreeding blanket options are selected for further design analysis and integration with a suitable power conversion subsystem

  7. Thermal and mechanical design of WITAMIR-I blanket

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.N.

    1980-10-01

    The design philosophy of WITAMIR-I, a Wisconsin Tandem Mirror Reactor design study, uses the experience obtained from our previous tokamak studies and combines it with the unique features of the tandem mirror to obtain an attractive design of a TM power reactor. It is aimed at maximizing the strengths of the tandem mirror while mitigating its weaknesses. The end product should be a safe, reliable, maintainable and a relatively economic power reactor. The general description of the reactor, the plasma calculations, the magnet design, the neutronic calculations and the maintenance considerations are presented elsewhere. This paper presents the blanket design of this reactor study

  8. Technical evaluation of major candidate blanket systems for fusion power reactor

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Seki, Masahiro; Minato, Akio

    1987-03-01

    The key functions required for tritium breeding blankets for a fusion power reactor are: (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li 2 O/H 2 O/Be, Mo-alloy/Li 2 O/He/Be, Mo-alloy/LiAlO 2 /He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. (author)

  9. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    Gohar, Y.; Baker, C.C.; Smith, D.L.

    1987-10-01

    The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the protection criteria required for different reactor components and dose equivalent after shutdown in the reactor hall. It is assumed that the blanket operation at commercial power reactor conditions and the proper temperature for power generation can be sacrificed to achieve the highest possible tritium breeding ratio with minimum additional research and developments and minimal impact on reactor design and operation. A set of blanket evaluation criteria has been used to compare possible blanket concepts. Six areas: performance, operating requirements, impact on reactor design and operation, safety and environmental impact, technology assessment, and cost have been defined for the evaluation process. A water-cooled blanket was developed to operate with a low temperature and pressure. The developed blanket contains a 24 cm of beryllium and 6 cm of solid breeder both with a 0.8 density factor. This blanket provides a local tritium breeding ratio of ∼2.0. The water coolant is isolated from the breeder material by several zones which eliminates the tritium buildup in the water by permeation and reduces the changes for water-breeder interaction. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. 12 refs., 13 tabs

  10. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    Noda, Kenji

    1998-03-01

    This report is the Proceedings of ''the Sixth International Workshop on Ceramic Breeder Blanket Interactions'' which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: 1) fabrication and characterization of ceramic breeders, 2) properties data for ceramic breeders, 3) tritium release characteristics, 4) modeling of tritium behavior, 5) irradiation effects on performance behavior, 6) blanket design and R and D requirements, 7) hydrogen behavior in materials, and 8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li 2 TiO 3 , tritium release behavior of Li 2 TiO 3 and Li 2 ZrO 3 including tritium diffusion, modeling of tritium release from Li 2 ZrO 3 in ITER condition, helium release behavior from Li 2 O, results of tritium release irradiation tests of Li 4 SiO 4 pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  11. Failure Modes and Effects Analysis on ITER DFLL-TBM system

    International Nuclear Information System (INIS)

    Hu Liqin; Yuan Run; Chen Hongli; Bai Yunqing

    2012-01-01

    As required for licensing process, accident analyses of International Thermonuclear Experimental Reactor (ITER) accounting for site specifications and design changes will be updated. Chinese Dual-Functional Lithium-Lead-Test Blanket Module (DFLL-TBM) system is a key safety-related component of ITER, its detailed safety analysis, which was designated to demonstrate the integrated technologies of both Helium single coolant (SLL) blanket and Helium-LiPb dual coolant (DLL) blanket, was performed. Failure Modes and Effects Analysis (FMEA) was applied to perform the safety analysis of DFLL-TBM. This study described the process of FMEA studies on DFLL-TBM system. All safety-related Postulated Initiating Events (PIEs) was identified. And a set of PIEs recommended to be taken into account in the further deterministic transient analyses were defined for both SLL and DLL blanket concepts separately.

  12. Activation analysis for ITER design options

    International Nuclear Information System (INIS)

    Attaya, H.

    1995-09-01

    This paper presents a summary of the activation analyses that have been performed for the shielding blanket (SS/water) and for the breeding blanket (Li/V) of ITER design options. The activation code RACC-P, which has been modified for pulsed operation, has been used in these calculations. The spatial distributions of the radioactive inventory, decay heat, biological hazard potential, and the contact dose were calculated for the two designs for different operation modes and targeted fluences. A one-dimensional toroidal cylindrical geometrical model has been utilized to determine the neutron fluxes in the two designs. The results are normalized for an inboard and outboard neutron wall loadings of 0.91 and 1.2 MW/m 2 respectively

  13. ITER technology R and D during the EDA

    International Nuclear Information System (INIS)

    Mizoguchi, T.

    2001-01-01

    A short overview of the ITER technology R and D achievements is presented. It includes R and D programme in the area of superconducting magnets, L-1 central solenoid model coil, L-2 toroidal field model coil, L-3 vacuum vessel sector, L-4 blanket module, L-5 divertor cassette, L-6 blanket and L-7 divertor remote handling systems. In addition to the seven large R and D projects, development of components for fuelling, pumping, tritium processing, heating/current drive, power supplies and plasma diagnostics, as well as safety-related R and D have significantly progressed

  14. Blanket/first wall challenges and required R&D on the pathway to DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, Mohamed, E-mail: abdou@fusion.ucla.edu; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-11-15

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  15. Blanket/first wall challenges and required R&D on the pathway to DEMO

    International Nuclear Information System (INIS)

    Abdou, Mohamed; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-01-01

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  16. Evaluation of steam as a potential coolant for nonbreeding blanket designs

    International Nuclear Information System (INIS)

    Stevens, H.C.; Misra, B.; Youngdahl, C.K.

    1978-01-01

    A steam-cooled nonbreeding blanket design has been developed as an evolution of the Argonne Experimental Power Reactor (EPR) studies. This blanket concept complete with maintenance considerations is to function at temperatures up to 650 0 C utilizing nickel-based alloys such as Inconel 625. Thermo-mechanical analyses were carried out in conjunction with thermal hydraulic analysis to determine coolant chennel arrangements that permit delivery of superheated steam at 500 0 C directly to a modern fossil plant-type turbine. A dual-cycle system combining a pressurized water circuit coupled with a superheated steam circuit can produce turbine plant conversion efficiencies approaching 41.5%

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  18. (D,T) Driven thorium hybrid blankets

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.; Khan, S.; Sahin, S.

    1983-01-01

    Recently, a project has started, with the aim to establish the neutronic performance and the basic design of an experimental fusionfission (hybrid) reactor facility, called AYMAN, in cylinderical geometry. The fusion reactor will have to be simulated by a (D,T) neutron generator. Fissile and fertile fuel will have to surround the neutron generator as a cylinderical blanket to simulate the boundary conditions of the hybrid blanket in a proper way. This geometry is consistent with Tandem Mirror Hybrid Blanket design and with most of the ICF blanket designs. A similar experimental installation will become operational around 1984 at the Swiss Federal Institute of Technology in Lausanne, Switzerland known under the project LOTUS. Due to the limited dimensions of the experimental cavity of the LOTUS-hybrid reactor, the LOTUS blankets have to be designed in plane geometry. Also, the bulky form of the Haefely neutron generator of the LOTUS facility obliges one to design a blanket in the plane geometry. This results in a vacuum left boundary conditions for the LOTUS blanket. The importance of a reflecting left boundary condition on the overall neutronic performance of a hybrid blanket has been analyzed in previous work in detail

  19. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

    1982-06-01

    The overall objective of this task is to investigate the extent to which the power density in the FED/INTOR breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow advanced reactor blanket modules to be tested on FED/INTOR under representative conditions

  20. Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

    International Nuclear Information System (INIS)

    Palermo, I.; Gómez-Ros, J.M.; Veredas, G.; Sanz, J.; Sedano, L.

    2012-01-01

    Highlights: ► Dual-Coolant He/Pb15.7Li breeding blanket for a DEMO fusion reactor is studied. ► An iterative process optimizes neutronic responses minimizing reactor dimension. ► A 3D toroidally symmetric geometry has been generated from the CAD model. ► Overall TBR values support the feasibility of the conceptual model considered. ► Power density in TF coils is below load limit for quenching. - Abstract: The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.

  1. Contact dose rates and relevant radioactive inventory in ITER TBM systems

    International Nuclear Information System (INIS)

    Zucchetti, M.; Guerrini, L.; Poitevin, Y.; Ricapito, I.; Zmitko, M.

    2011-01-01

    The determination of the radioactive inventory and of the contact dose rates in the different ITER Test Blanket Modules systems is carried out, both for Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, already available for both the blanket modules. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated.

  2. Contact dose rates and relevant radioactive inventory in ITER TBM systems

    Energy Technology Data Exchange (ETDEWEB)

    Zucchetti, M., E-mail: massimo.zucchetti@polito.it [EURATOM/ENEA Fusion Association Politecnico di Torino, Torino (Italy); Guerrini, L., E-mail: Laurent.Guerrini@f4e.europa.eu [Fusion for Energy, ITER Department, Test Blanket Modules Group, Barcelona (Spain); Poitevin, Y.; Ricapito, I.; Zmitko, M. [Fusion for Energy, ITER Department, Test Blanket Modules Group, Barcelona (Spain)

    2011-10-15

    The determination of the radioactive inventory and of the contact dose rates in the different ITER Test Blanket Modules systems is carried out, both for Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, already available for both the blanket modules. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated.

  3. Lithium test module on ITER: engineering design of the tritium recovery system

    International Nuclear Information System (INIS)

    Finn, P.A.

    1988-01-01

    The design presented is an overview of the tritium recovery system for a lithium module on an ITER type reactor. The design of a tritium recovery system for larger blanket units, sectors, etc. could use the information developed in this report. A goal of this design was to ensure that a reliable, integrated performance of the tritium recovery system could be demonstrated. An equally important goal was to measure and account for the tritium in the liquid lithium blanket module and its recovery system in order to validate the operation of the blanket module

  4. Draining and drying process development of the Tokamak Cooling Water System of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seokho, E-mail: kims@ornl.gov [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Van Hove, Walter; Ferrada, Juan [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Di Maio, Pietro Alessandro [University of Palermo, Viale delle Scienze, Palermo 90128 (Italy); Felde, David [Reactor and Nuclear Systems Division, ORNL, Oak Ridge, TN (United States); Raphael, Mitteau; Dell’Orco, Giovanni [ITER Organization, 13067 St Paul Lez Durance (France); Berry, Jan [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    2016-11-01

    Highlights: • A thermal-hydraulic model using RELAP was developed for the ITER FW/BLK modules to determine design parameters for the nitrogen blowout flow rate and pressure. • The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will sufficiently evacuate the water in blankets. • A limited validation study indicates that the analysis yields less conservative results to compare against data collected from experiments. Therefore, the designed blow out flow of the drying system was selected with a large margin above the measured values to ensure the blow out operation. - Abstract: The ITER Organization (IO) developed a thermal-hydraulic (TH) model of the complex first wall and blanket (FW/BLK) cooling channels to determine gas flow rate and pressure required to effectively blow out the water in the FW/BLK. In addition, US ITER conducted experiments for selected geometries of FW/BLK flow channels to predict the blowout parameters. The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will ensure substantial evacuation of the water in blankets with just a few percent remaining in the blanket flow channels. A limited validation study indicates that the analysis yields less conservative results to compare against data collected from experiments. Therefore, the designed blow out flow of the drying system was selected with a large margin above the measured values to ensure the blow out operation.

  5. ITER fuel cycle systems layout

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-10-01

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

  6. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    2009-01-01

    Selecting relevant features is in demand when a large data set is of interest in a classification task. It produces a tractable number of features that are sufficient and possibly improve the classification performance. This paper studies a statistical method of Markov blanket induction algorithm...... for filtering features and then applies a classifier using the Markov blanket predictors. The Markov blanket contains a minimal subset of relevant features that yields optimal classification performance. We experimentally demonstrate the improved performance of several classifiers using a Markov blanket...... induction as a feature selection method. In addition, we point out an important assumption behind the Markov blanket induction algorithm and show its effect on the classification performance....

  7. Blanket materials for DT fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1981-01-01

    This paper presents an overview of the critical materials issues that must be considered in the development of a tritium breeding blanket for a tokamak fusion reactor that operates on the D-T-Li fuel cycle. The primary requirements of the blanket system are identified and the important criteria that must be considered in the development of blanket technology are summarized. The candidate materials are listed for the different blanket components, e.g., breeder, coolant, structure and neutron multiplier. Three blanket concepts that appear to offer the most potential are: (1) liquid-metal breeder/coolant, (2) liquid-metal breeder/separate coolant, and (3) solid breeder/separate coolant. The major uncertainties associated with each of the design concepts are discussed and the key materials R and D requirements for each concept are identified

  8. Status of the EU test blanket systems safety studies

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-01-01

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  9. Status of the EU test blanket systems safety studies

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-10-15

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  10. Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

    Science.gov (United States)

    Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

    2017-08-01

    The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

  11. 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

  12. 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

  13. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

    Diegele, E.

    2009-01-01

    Materials development in nuclear fusion for in-vessel components, i.e. for breeder blankets and divertors, has a history of more than two decades. It is the specific in-service and loading conditions and the consequentially required properties in combination with safety standards and social-economic demands that create a unique set of specifications. Objectives of Fusion for Energy (F4E) include: 1) To provide Europe's contribution to the ITER international fusion energy project; 2) To implement the Broader Approach agreement between Euratom and Japan; 3) To prepare for the construction and demonstration of fusion reactors (DEMO). Consequently, activities in F4E focus on structural materials for the first generations of breeder blankets, i.e. ITER Test Blanket Modules (TBM) and DEMO, whereas a Fusion Materials Topical Group implemented under EFDA coordinates R and D on physically based modelling of irradiation effects and R and D in the longer term (new and /or higher risk materials). The paper focuses on martensitic-ferritic steels and (i) reviews briefly the challenges and the rationales for the decisions taken in the past, (ii) analyses the status of the main activities of development and qualification, (iii) indicates unresolved issues, and (iv) outlines future strategies and needs and their implications. Due to the exposure to intense high energy neutron flux, the main issue for breeder materials is high radiation resistance. The First Wall of a breeder blanket should survive 3-5 full power years or, respectively in terms of irradiation damage, typically 50-70 dpa for DEMO and double figures for a power plant. Even though the objective is to have the materials and key fabrication technologies needed for DEMO fully developed and qualified within the next two decades, a major part of the task has to be completed much earlier. Tritium breeding test blanket modules will be installed in ITER with the objective to test DEMO relevant technologies in fusion

  14. 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

  15. A solid-breeder blanket and power conversion system for the Mirror Advanced Reactor Study (MARS)

    International Nuclear Information System (INIS)

    Bullis, R.; Clarkson, I.

    1983-01-01

    A solid-breeder blanket has been designed for a commercial fusion power reactor based on the tandem mirror concept (MARS). The design utilizes lithium oxide, cooled by helium which powers a conventional steam electric generating cycle. Maintenance and fabricability considerations led to a modular configuration 6 meters long which incorporates two magnets, shield, blanket and first wall. The modules are arranged to form the 150 meter long reactor central cell. Ferritic steel is used for the module primary structure. The lithium oxide is contained in thin-walled vanadium alloy tubes. A tritium breeding ratio of 1.25 and energy multiplication of 1.1 is predicted. The blanket design appears feasible with only a modest advance in current technology

  16. 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.

  17. ITER Equatorial Port plug engineering: Design and remote handling activities supported by Virtual Reality tools

    International Nuclear Information System (INIS)

    Keller, Delphine; Dechelle, Christian; Doceul, Louis; Madeleine, Sylvain; Martins, Jean Pierre; Measson, Yvan; Patterlini, Jean Claude; Wagrez, Julien

    2011-01-01

    In the context of ITER, CEA/IRFM has participated to the design and integration of several components in the Equatorial Port plug region. Particularly, in the framework of the grant F4E-2008-GRT-09-PNS-TBM, CEA/IRFM has contributed to the test blanket module system (TBS) design and robot access feasibility study in the Port Cell. Simulations of the maintenance procedure were studied and fully integrated to the design process, enabling to provide space reservation for human and robotic access. For this mean, CEA/IRFM has used a CEA LIST Virtual Reality simulation software directly integrated to the Solidworks CAD software. The feasibility to connect/dis-connect the pipes in front of the Bioshield by a set of potential standard industrial arms was demonstrated. Aiming to give more realism to maintenance scenario and CAD models, CEA IRFM has decided to build a Virtual Reality platform in the institute, integrated to the design office. With the expertise of CEA LIST, this platform aims to provide the nearest possible links between design and remote handling needs. This paper presents the outcome of the robot access study and discusses about the Virtual Reality tools that are being developed for these applications.

  18. ITER Equatorial Port plug engineering: Design and remote handling activities supported by Virtual Reality tools

    Energy Technology Data Exchange (ETDEWEB)

    Keller, Delphine, E-mail: delphine.keller@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Dechelle, Christian; Doceul, Louis; Madeleine, Sylvain; Martins, Jean Pierre [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Measson, Yvan [CEA, LIST, Interactive Robotics Unit, 18 route du Panorama, BP6, F-92265 Fontenay Aux Roses (France); Patterlini, Jean Claude; Wagrez, Julien [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France)

    2011-10-15

    In the context of ITER, CEA/IRFM has participated to the design and integration of several components in the Equatorial Port plug region. Particularly, in the framework of the grant F4E-2008-GRT-09-PNS-TBM, CEA/IRFM has contributed to the test blanket module system (TBS) design and robot access feasibility study in the Port Cell. Simulations of the maintenance procedure were studied and fully integrated to the design process, enabling to provide space reservation for human and robotic access. For this mean, CEA/IRFM has used a CEA LIST Virtual Reality simulation software directly integrated to the Solidworks CAD software. The feasibility to connect/dis-connect the pipes in front of the Bioshield by a set of potential standard industrial arms was demonstrated. Aiming to give more realism to maintenance scenario and CAD models, CEA IRFM has decided to build a Virtual Reality platform in the institute, integrated to the design office. With the expertise of CEA LIST, this platform aims to provide the nearest possible links between design and remote handling needs. This paper presents the outcome of the robot access study and discusses about the Virtual Reality tools that are being developed for these applications.

  19. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

    1993-11-01

    The ARIES-IV Nested Shell Blanket (NSB) Design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the diverter design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of I (LSA-1), which indicates an inherently safe design

  20. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

    1999-07-01

    The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

  1. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

    2017-01-15

    Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

  2. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    International Nuclear Information System (INIS)

    WONG, CPC; MALANG, S; NISHIO, S; RAFFRAY, R; SAGARA, S

    2002-01-01

    OAK A271 ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS. First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability