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Sample records for blanket remote handling

  1. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  3. Dust removal experiments for ITER blanket remote handling system

    International Nuclear Information System (INIS)

    To reduce maintenance workers' dose rate caused by activated dust adhering to the ITER blanket remote handling system (BRHS), dust must be removed from BRHS surfaces. Dust that adheres to the top surface of the BRHS rail from cyclic loading of the vehicle manipulator is considered to be the most difficult dust to remove. Dust removal experiments were conducted to simulate the materials, conditions, and cyclic loading of actual BRHS operations. The tungsten powder used to simulate the dust was squashed, and the area of contact by cyclic load was increased, but the powder was not embedded into the matrix. The increase in the area of contact increased the total intermolecular force between a tungsten particle and the surface, which was considered the main force adhering dust to the test piece surface. A combination of dust removal methods, including vacuum cleaning and brushing, was applied to the simulated dust that adhered to the test pieces. The results showed that vacuum cleaning is effective in removing dust from the non-cyclic loaded surface. The combined methods were highly efficient in removing the dust that strongly adhered to the rail surface. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  6. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  9. In-vessel remote handling machine for blanket replacement in the demo fusion reactor

    International Nuclear Information System (INIS)

    The paper presents the current state of investigations concerning the adaptation of the ITER in-vessel remote handling system of 1998 to DEMO conditions. The outline of the concept is the following: a rail is built up in the middle of the vessel along the major radius forming a full circle. It is supported from the four equatorial ports by long radial arms connected perpendicularly to the rail. On the rail four manipulators with telescopic arms are operating each being responsible for a 90 deg section of the rail. Within their section the manipulators are capable of reaching and removing every element, and can manipulate 10 t elements at 3,5 m distance with great precision. Element exchange will take place through the lower section of the ports. Great advantage of the system is that it is only supported from the ports, thus the maintenance of the divertor and blanket can be planned independently. For this reason the system is preferred for DEMO, but there are challenges to face, and they come from the large and heavy elements having to be inserted through the ports of limited size, the complicated installation process and the need for precision. The results of this work indicate that this adaptation can be done, although more investigation is necessary regarding the manipulator design (author)

  10. Remote-handling concept for target/blanket modules in the accelerator production of tritium

    International Nuclear Information System (INIS)

    The accelerator production of tritium (APT) has been proposed as the source of tritium for the United States in the next century. The APT will accelerate protons that will strike replaceable tungsten target modules. The tungsten target modules generate neutrons that pass through blanket modules and other modules where He gas is turned into tritium. The target and blanket modules are predicted to require replacement every 1 to 10 yr, depending on their location. The target modules may weigh as much as 78.8 tonnes (85 t) each. All of the modules will be contained in a target/blanket vessel, which is in a shielded facility. The spent modules will be very radioactive so that remote replacement is required. A proposed concept is to use a remotely operated bridge crane and a remotely operated, bridge-mounted manipulator to perform the entire replacement operation. This will require removing/replacing the vessel lid, installing/removing temporary water cooling, closing/opening valves on manifolds and modules, draining of jumpers, removing/replacing jumpers, removing/replacing shielding keys, and removing/replacing the modules. This application is unique because of the size and weight of the modules, the precision required, the type of connectors required, and the complexity of the entire operation. A three-dimensional simulation of the entire module replacement operation has been developed to help understand, communicate, and refine the concepts

  11. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

    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)

  12. Progress of R and D and design of blanket remote handling equipment for ITER

    International Nuclear Information System (INIS)

    The design of in-vessel transporter (IVT) including vehicle manipulator has been updated according to the design changes such as blanket segmentation and structure, taking account of the interface between modules and vehicle manipulator. In particular, the updated design of the vehicle manipulator and rail has been carried out because of collision avoidance between modules and vehicle manipulator. According to the updated design, the vehicle manipulator has been reduced by about 30% in weight, compared with the reference design. In parallel with design activities, the R and D to clarify the specifications of the IVT design in detail is also performed, i.e., simulation system to provide the visual information during maintenance, dry lubricant to prevent the lubricant oil from spreading in the vacuum vessel (VV). The rail connection and cable handling in the transfer cask, which are critical issues for IVT system, are under preparation of the demonstration tests to finalize the design of the IVT system. Connection of the rail joint and cable handling test facilities are planned and under fabrication now. These test facility will be installed by the end of March 2008, and the performance tests will be carried out from April 2008

  13. 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)

    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)

  14. Remote handling at LAMPF

    International Nuclear Information System (INIS)

    Experimental area A at the Clinton P. Anderson Meson Physics Facility (LAMPF) encompasses a large area. Presently there are four experimental target cells along the main proton beam line that have become highly radioactive, thus dictating that all maintenance be performed remotely. The Monitor remote handling system was developed to perform in situ maintenance at any location within area A. Due to the complexity of experimental systems and confined space, conventional remote handling methods based upon hot cell and/or hot bay concepts are not workable. Contrary to conventional remote handling which require special tooling for each specifically planned operation, the Monitor concept is aimed at providing a totally flexible system capable of remotely performing general mechanical and electrical maintenance operations using standard tools. The Monitor system is described

  15. 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.

  16. Advanced technologies for remote handling

    International Nuclear Information System (INIS)

    Master slave manipulators (MSMs), in-cell cranes and power manipulators are the general-purpose remote handling tools used in nuclear industry. In-cell cranes and power manipulators can handle heavy objects; whereas MSMs can handle objects with precision and dexterity. The department had identified the importance of indigenising these technologies and developed a variety of mechanical MSMs and Servo Manipulators. This paper traces the history and evolution of these technologies. It also mentions about the telepresence technologies that are set to transform the operator's experience of manipulation by bringing in visual, haptic and aural immersion in the slave environment. (author)

  17. Remote handling equipment for SNS

    International Nuclear Information System (INIS)

    This report gives information on the areas of the SNS, facility which become highly radioactive preventing hands-on maintenance. Levels of activity are sufficiently high in the Target Station Area of the SNS, especially under fault conditions, to warrant reactor technology to be used in the design of the water, drainage and ventilation systems. These problems, together with the type of remote handling equipment required in the SNS, are discussed

  18. Remote Handling System for Ignitor^*

    Science.gov (United States)

    Galbiati, L.; Bianchi, A.; Lucca, F.; Coppi, B.

    2005-10-01

    Since access in Ignitor is through the limited width of the equatorial ports, the use of remote handling (RH) technology for any in-vessel intervention is required, even before the vessel becomes activated. In particular, the first wall of Ignitor, which is made of TZM (Molybdenum) tiles mounted on Inconel tile-carriers covering the entire plasma chamber, has been designed to be installed and replaced entirely by the RH system. The presence of radiation screens inside the cryostat and around the ports ensure a sufficiently low level of activation around the machine to avoid the need of ex-vessel RH techniques. The in-vessel RH system is based on two transporters carrying an articulated boom with end-effectors, supported by a movable structure over a transport system that can be lifted and set in position adjacent to two opposite horizontal ports. The design of the in-vessel RH system, of the boom and its enclosure, and of the most significant end-effectors (welding and cutting tools, and tools for the removal and handling of tile carriers) has been completed. A series of other dedicated tools for installation and maintainances of diagnostics components, of the RF antennas, vacuum cleaners, tools for general inspection and metrology are included in the design. ^*Sponsored in part by ENEA of Italy and by the U.S. DOE.

  19. Development of remote handling tools and equipment

    International Nuclear Information System (INIS)

    The remote handling (RH) tools and equipment development in ITER focuses mainly on the welding and cutting technique, weld inspection and double-seal door which are essential factors in the replacement of in-vessel components such as divertor and blanket. The conceptual design of these RH tools and equipment has been defined through ITER engineering design activity (EDA). Similarly, elementary R and D of the RH tools and equipment have been extensively performed to accumulate a technological data base for process and performance qualification. Based on this data, fabrications of full-scale RH tools and equipment are under progress. A prototypical bore tool for pipe welding and cutting has already been fabricated and is currently undergoing integrated performance tests. This paper describes the design outline of the RH tools and equipment related to in-vessel components maintenance, and highlights the current status of RH tools and equipment development by the Japan Home Team as an ITER R and D program. This paper also includes an outline of insulation joint and quick-pipe connector development, which has also been conducted through the ITER R and D program in order to standardize RH operations and components. (author)

  20. Remote Handling behind port plug in ITER

    International Nuclear Information System (INIS)

    Different Test Blanket Modules (TBM) will be used in succession in the same equatorial ports of ITER. The remote handling operations for connection/disconnection of an interface between the port plug of the EU-HCPB-TBM and the port cell equipment are investigated with the goal to reach a quick and simple TBM exchange procedure. This paper describes the operations and systems which are required for connection of the TBM to its supply lines at this interface. The interface is located inside the free space of the port plug flange between the port plug shield and the bioshield of the port cell behind. The approach of the operation place is only available through a narrow gate in the bioshield opened temporarily during maintenance periods. This gate limits the dimensions of the whole system and its tools. The current design of the EU-HCPB-TBM foresees up to 9 supply lines which have to be connected inside the free space of one half of the port plug flange. The connection operations require positioning and adjustment of the tools for each pipe separately. Despite the strict circumstances it is still possible to find such an industrial jointed-arm robot with sufficient payload, which can penetrate into the working area. A mechanical system is necessary to move the robot from its storing place in the hot cell to the port plug on 6 m distance. Each operation requires different end-of-arm tools. The most special one is a pipe positioner tool, which can position and pull the pipe ends to each other and align the tool before welding and hold them in proper position during the welding process. Weld seams can be made by orbital welding tool. The pipe positioner tool has to provide place for welding tool. Using of inbore tool is impossible because pipes have no open ends where the tool could leave it. Orbital tool must be modified to meet requirements of remote handling because it is designed for human handling. The coolant is helium, so for eliminating the leak of helium it is

  1. Apparatus for remotely handling components

    Science.gov (United States)

    Szkrybalo, Gregory A.; Griffin, Donald L.

    1994-01-01

    The inventive apparatus for remotely handling bar-like components which define a longitudinal direction includes a gripper mechanism for gripping the component including first and second gripper members longitudinally fixedly spaced from each other and oriented parallel to each other in planes transverse to the longitudinal direction. Each gripper member includes a jaw having at least one V-groove with opposing surfaces intersecting at a base and extending radially relative to the longitudinal direction for receiving the component in an open end between the opposing surfaces. The V-grooves on the jaw plate of the first and second gripper members are aligned in the longitudinal direction to support the component in the first and second gripper members. A jaw is rotatably mounted on and a part of each of the first and second gripper members for selectively assuming a retracted mode in which the open end of the V-groove is unobstructed and active mode in which the jaw spans the open end of the V-groove in the first and second gripper members. The jaw has a locking surface for contacting the component in the active mode to secure the component between the locking surface of the jaw and the opposing surfaces of the V-groove. The locking surface has a plurality of stepped portions, each defining a progressively decreasing radial distance between the base of the V-groove and the stepped portion opposing the base to accommodate varying sizes of components.

  2. Hot Laboratories and Remote Handling

    International Nuclear Information System (INIS)

    The Opening talk of the workshop 'Hot Laboratories and Remote Handling' was given by Marin Ciocanescu with the communication 'Overview of R and D Program in Romanian Institute for Nuclear Research'. The works of the meeting were structured into three sections addressing the following items: Session 1. Hot cell facilities: Infrastructure, Refurbishment, Decommissioning; Session 2. Waste, transport, safety and remote handling issues; Session 3. Post-Irradiation examination techniques. In the frame of Section 1 the communication 'Overview of hot cell facilities in South Africa' by Wouter Klopper, Willie van Greunen et al, was presented. In the framework of the second session there were given the following four communications: 'The irradiated elements cell at PHENIX' by Laurent Breton et al., 'Development of remote equipment for DUPIC fuel fabrication at KAERI', by Jung Won Lee et al., 'Aspects of working with manipulators and small samples in an αβγ-box, by Robert Zubler et al., and 'The GIOCONDA experience of the Joint Research Centre Ispra: analysis of the experimental assemblies finalized to their safe recovery and dismantling', by Roberto Covini. Finally, in the framework of the third section the following five communications were presented: 'PIE of a CANDU fuel element irradiated for a load following test in the INR TRIGA reactor' by Marcel Parvan et al., 'Adaptation of the pole figure measurement to the irradiated items from zirconium alloys' by Yury Goncharenko et al., 'Fuel rod profilometry with a laser scan micrometer' by Daniel Kuster et al., 'Raman spectroscopy, a new facility at LECI laboratory to investigate neutron damage in irradiated materials' by Lionel Gosmain et al., and 'Analysis of complex nuclear materials with the PSI shielded analytical instruments' by Didier Gavillet. In addition, eleven more presentations were given as posters. Their titles were: 'Presentation of CETAMA activities (CEA analytic group)' by Alain Hanssens et al. 'Analysis of

  3. Remote handling maintenance of ITER

    International Nuclear Information System (INIS)

    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)

  4. Remote handling maintenance of ITER

    International Nuclear Information System (INIS)

    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 feasibility in principle of the remote maintenance procedures and to develop the required equipment and tools, were operational at the end of the Engineering Design Activities phase. The initial results are highly encouraging: major remote equipment deployment and component replacement operations have been successfully demonstrated. (author)

  5. Mechanical characteristics and position control of vehicle/manipulator for ITER blanket remote maintenance

    International Nuclear Information System (INIS)

    In International Thermonuclear Experimental Reactor (ITER), blanket maintenance requires the 4-tonne module handling with high positioning accuracy of ±2 mm. In order to meet this requirement, it is essential to suppress the dynamic deflection and vibration of the remote handling equipment due to sudden transfer of the module weight from/to the back-plate supports to/from the equipment itself during installation and removal. A new control scheme was proposed and tested so as to suppress the dynamic behaviors. As a result, the dynamic deflection of the rail and the acceleration of the manipulator were successfully decreased to nearly zero. Based on the test results, the proposed control scheme was concluded to be effective so as to suppress this kind of dynamic effect during heavy component handling

  6. Canadian capabilities in fusion fuels technology and remote handling

    International Nuclear Information System (INIS)

    This report describes Canadian expertise in fusion fuels technology and remote handling. The Canadian Fusion Fuels Technology Project (CFFTP) was established and is funded by the Canadian government, the province of Ontario and Ontario Hydro to focus on the technology necessary to produce and manage the tritium and deuterium fuels to be used in fusion power reactors. Its activities are divided amongst three responsibility areas, namely, the development of blanket, first wall, reactor exhaust and fuel processing systems, the development of safe and reliable operating procedures for fusion facilities, and, finally, the application of these developments to specific projects such as tritium laboratories. CFFTP also hopes to utilize and adapt Canadian developments in an international sense, by, for instance, offering training courses to the international tritium community. Tritium management expertise is widely available in Canada because tritium is a byproduct of the routine operation of CANDU reactors. Expertise in remote handling is another byproduct of research and development of of CANDU facilities. In addition to describing the remote handling technology developed in Canada, this report contains a brief description of the Canadian tritium laboratories, storage beds and extraction plants as well as a discussion of tritium monitors and equipment developed in support of the CANDU reactor and fusion programs. Appendix A lists Canadian manufacturers of tritium equipment and Appendix B describes some of the projects performed by CFFTP for offshore clients

  7. Remote-handled transuranic system assessment appendices. Volume 2

    International Nuclear Information System (INIS)

    Volume 2 of this report contains six appendices to the report: Inventory and generation of remote-handled transuranic waste; Remote-handled transuranic waste site storage; Characterization of remote-handled transuranic waste; RH-TRU waste treatment alternatives system analysis; Packaging and transportation study; and Remote-handled transuranic waste disposal alternatives

  8. Reliability of robotics: an overview with identification of specific aspects related to remote handling in fusion machines

    International Nuclear Information System (INIS)

    This paper describes the derivation of a reliability and safety analysis methodology, and its application in a case study of a handling device for the blankets that constitute the fusion reactor torus. This is taken as an opportunity to identify the particularities of reliability and safety when related to remote handling in fusion machines. (orig.)

  9. A Perspective on Remote Handling Operations and Human Machine Interface for Remote Handling in Fusion

    International Nuclear Information System (INIS)

    A large-scale fusion device presents many challenges to the remote handling operations team. This paper is based on unique operational experience at JET and gives a perspective on remote handling task development, logistics and resource management, as well as command, control and human-machine interface systems. Remote operations require an accurate perception of a dynamic environment, ideally providing the operators with the same unrestricted knowledge of the task scene as would be available if they were actually at the remote work location. Traditional camera based systems suffer from a limited number of viewpoints and also degrade quickly when exposed to high radiation. Virtual Reality and Augmented Reality software offer great assistance. The remote handling system required to maintain a tokamak requires a large number of different and complex pieces of equipment coordinating to perform a large array of tasks. The demands on the operator's skill in performing the tasks can escalate to a point where the efficiency and safety of operations are compromised. An operations guidance system designed to facilitate the planning, development, validation and execution of remote handling procedures is essential. Automatic planning of motion trajectories of remote handling equipment and the remote transfer of heavy loads will be routine and need to be reliable. This paper discusses the solutions developed at JET in these areas and also the trends in management and presentation of operational data as well as command, control and HMI technology development offering the potential to greatly assist remote handling in future fusion machines. (author)

  10. Remote technologies for handling spent fuel

    International Nuclear Information System (INIS)

    The nuclear programme in India involves building and operating power and research reactors, production and use of isotopes, fabrication of reactor fuel, reprocessing of irradiated fuel, recovery of plutonium and uranium-233, fabrication of fuel containing plutonium-239, uranium-233, post-irradiation examination of fuel and hardware and handling solid and liquid radioactive wastes. Fuel that could be termed 'spent' in thermal reactors is a source for second generation fuel (plutonium and uranium-233). Therefore, it is only logical to extend remote techniques beyond handling fuel from thermal reactors to fuel from fast reactors, post-irradiation examination etc. Fabrication of fuel containing plutonium and uranium-233 poses challenges in view of restriction on human exposure to radiation. Hence, automation will serve as a step towards remotisation. Automated systems, both rigid and flexible (using robots) need to be developed and implemented. Accounting of fissile material handled by robots in local area networks with appropriate access codes will be possible. While dealing with all these activities, it is essential to pay attention to maintenance and repair of the facilities. Remote techniques are essential here. There are a number of commonalities in these requirements and so development of modularized subsystems, and integration of different configurations should receive attention. On a long-term basis, activities like decontamination, decommissioning of facilities and handling of waste generated have to be addressed. While robotized remote systems have to be designed for existing facilities, future designs of facilities should take into account total operation with robotic remote systems. (author)

  11. Remote handling systems for the Pride application

    International Nuclear Information System (INIS)

    In this paper is described the development of remote handling systems for use in the pyro processing technology development. Remote handling systems mainly include a BDSM (Bridge transported Dual arm Servo-Manipulator) and a simulator, all of which will be applied to the Pride (Pyro process integrated inactive demonstration facility) that is under construction at KAERI. BDMS that will traverse the length of the ceiling is designed to have two pairs of master-slave manipulators of which each pair of master-slave manipulators has a kinematic similarity and a force reflection. A simulator is also designed to provide an efficient means for simulating and verifying the conceptual design, developments, arrangements, and rehearsal of the pyro processing equipment and relevant devices from the viewpoint of remote operation and maintenance. In our research is presented activities and progress made in developing remote handling systems to be used for the remote operation and maintenance of the pyro processing equipment and relevant devices in the Pride. (Author)

  12. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Park, B. S.; Yoon, J. S.; Hong, H. D. (and others)

    2007-02-15

    In this research, the remote handling technology was developed for the ACP application. The ACP gives a possible solution to reduce the rapidly cumulative amount of spent fuels generated from the nuclear power plants in Korea. The remote technologies developed in this work are a slitting device, a voloxidizer, a modified telescopic servo manipulator and a digital mock-up. A slitting device was developed to declad the spent fuel rod-cuts and collect the spent fuel UO{sub 2} pellets. A voloxidizer was developed to convert the spent fuel UO{sub 2} pellets obtained from the slitting process in to U{sub 3}O{sub 8} powder. Experiments were performed to test the capabilities and remote operation of the developed slitting device and voloxidizer by using simulated rod-cuts and fuel in the ACP hot cell. A telescopic servo manipulator was redesigned and manufactured improving the structure of the prototype. This servo manipulator was installed in the ACP hot cell, and the target module for maintenance of the process equipment was selected. The optimal procedures for remote operation were made through the maintenance tests by using the servo manipulator. The ACP digital mockup in a virtual environment was established to secure a reliability and safety of remote operation and maintenance. The simulation for the remote operation and maintenance was implemented and the operability was analyzed. A digital mockup about the preliminary conceptual design of an enginnering-scale ACP was established, and an analysis about a scale of facility and remote handling was accomplished. The real-time diagnostic technique was developed to detect the possible fault accidents of the slitting device. An assessment of radiation effect for various sensors was also conducted in the radiation environment.

  13. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    In this research, the remote handling technology was developed for the ACP application. The ACP gives a possible solution to reduce the rapidly cumulative amount of spent fuels generated from the nuclear power plants in Korea. The remote technologies developed in this work are a slitting device, a voloxidizer, a modified telescopic servo manipulator and a digital mock-up. A slitting device was developed to declad the spent fuel rod-cuts and collect the spent fuel UO2 pellets. A voloxidizer was developed to convert the spent fuel UO2 pellets obtained from the slitting process in to U3O8 powder. Experiments were performed to test the capabilities and remote operation of the developed slitting device and voloxidizer by using simulated rod-cuts and fuel in the ACP hot cell. A telescopic servo manipulator was redesigned and manufactured improving the structure of the prototype. This servo manipulator was installed in the ACP hot cell, and the target module for maintenance of the process equipment was selected. The optimal procedures for remote operation were made through the maintenance tests by using the servo manipulator. The ACP digital mockup in a virtual environment was established to secure a reliability and safety of remote operation and maintenance. The simulation for the remote operation and maintenance was implemented and the operability was analyzed. A digital mockup about the preliminary conceptual design of an enginnering-scale ACP was established, and an analysis about a scale of facility and remote handling was accomplished. The real-time diagnostic technique was developed to detect the possible fault accidents of the slitting device. An assessment of radiation effect for various sensors was also conducted in the radiation environment

  14. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H

    1997-12-01

    Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  15. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  16. ITER - TVPS remote handling critical design issues

    International Nuclear Information System (INIS)

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

  17. Solution for remote handling in accelerator installations

    International Nuclear Information System (INIS)

    A description is given of a remote-handling system designed for the Los Alamos Clinton P. Anderson Meson Physics Facility (LAMPF), versatile enough to be used in a variety of situations found around particle accelerators. The system consists of a bilateral (force-reflecting) servomanipulator installed on an articulated hydraulic boom. The boom also carries the necessary tools and observation devices. The whole slave unit can be moved by crane or truck to the area of operation. A control cable connects the slave unit with the control station, located at a safe distance in a trailer. Various stages of development as well as some operating experience are discussed

  18. Evaluating ITER remote handling middleware concepts

    International Nuclear Information System (INIS)

    Highlights: ► Remote Handling Study Centre: middleware system setup and modules built. ► Aligning to ITER RH Control System Layout: prototype of database, VR and simulator. ► OpenSplice DDS, ZeroC ICE messaging and object oriented middlewares reviewed. ► Windows network latency found problematic for semi-realtime control over the network. -- Abstract: Remote maintenance activities in ITER will be performed by a unique set of hardware systems, supported by an extensive software kit. A layer of middleware will manage and control a complex set of interconnections between teams of operators, hardware devices in various operating theatres, and databases managing tool and task logistics. The middleware is driven by constraints on amounts and timing of data like real-time control loops, camera images, and database access. The Remote Handling Study Centre (RHSC), located at FOM institute DIFFER, has a 4-operator work cell in an ITER relevant RH Control Room setup which connects to a virtual hot cell back-end. The centre is developing and testing flexible integration of the Control Room components, resulting in proof-of-concept tests of this middleware layer. SW components studied include generic human-machine interface software, a prototype of a RH operations management system, and a distributed virtual reality system supporting multi-screen, multi-actor, and multiple independent views. Real-time rigid body dynamics and contact interaction simulation software supports simulation of structural deformation, “augmented reality” operations and operator training. The paper presents generic requirements and conceptual design of middleware components and Operations Management System in the context of a RH Control Room work cell. The simulation software is analyzed for real-time performance and it is argued that it is critical for middleware to have complete control over the physical network to be able to guarantee bandwidth and latency to the components

  19. Highly active vitrification plant remote handling operational experience and improvements

    International Nuclear Information System (INIS)

    Operational experience and technological innovation in the area of remote handling is described for the Sellafield Waste Vitrification Plant (WVP). This plant turns Highly Active Liquid Wastes (HALW) into radioactively immobile, solid forms. The technology needed for remote handling of HALWs, such as ejectors and power fluidics is described as is the mechanical handling needed after the vitrification process. Key features of WVP are described, such as the in-cell cranes, master-slave manipulators and swabbing robots. The severity of the in-cell environment has highlighted the need for innovation in the remote handling equipment and these changes are also described. (UK)

  20. DEMO hot cell and ex-vessel remote handling

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Justin, E-mail: justin.thomas@ccfe.ac.uk [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Loving, Antony [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Bachmann, Christian; Harman, Jon [EFDA, PPP and T, Boltzmannstraße 2, 85748 Garching (Germany)

    2013-10-15

    Highlights: ► Overview of current DEMO maintenance concepts. ► Comparison of current dextrous remote handling technologies and their rebalance to DEMO. ► Presentation of some ideas to improve the productivity and reliability of the DEMO ex-vessel transport system. ► A description of the size and type facilities that might be required in the DEMO hot cell. ► Identification of some areas that need to be developed further to meet the requirements of DEMO. -- Abstract: In Europe the work on the specification and design of a demonstration power plant (DEMO) is being carried out by EFDA in the power plant physics and technology (PPP and T) programme. DEMO will take fusion from experimental research into showing the potential for commercial power generation. During the fusion reaction, components in the tokamak become highly activated. The estimated dose rate levels after shutdown (zero decay time) due to 60 dpa accumulation in steel (blanket) and 30 dpa (divertor) are 13.1–17.4 kGy/h (blanket); 8.8–11.6 kGy/h (divertor) [1], much higher than those to be encountered at ITER. Upon removal from the tokamak, components would be transported to the hot cell facility with attention to minimizing the spread of activated dust and tritium contamination. It is proposed to use a sealed cask of ∼20 tonnes, running on air castors with 50% lifting capability redundancy. Due to the number and complexity of the routes taken by this transporter it would have to be an un-tethered semi-autonomous system. This poses some technical challenges, including providing sufficient battery capacity, reliable guidance and a fail safe un-tethered control system. The mass of the components being moved is assumed here to range from a few tonnes to in excess of one hundred tonnes. Before the removed in-vessel components can be processed in the hot cell, they would require a period of cooling, approximately 2.5 years, to allow dose rate and decay heating to reduce. This reduces the decay

  1. Beginnings of remote handling at the RAL Spallation Neutron Source

    International Nuclear Information System (INIS)

    Expenditure of funds and resources for remote maintenance systems traditionally are delayed until late in an accelerator's development. However, simple remote-surveillance equipment can be included early in facility planning to set the stage for future remote-handling needs and to identify appropriate personnel. Some basic equipment developed in the UK at the Spallation Neutron Source (SNS) that serves this function and that has been used to monitor beam loss during commissioning is described. A photograph of this equipment, positioned over the extractor septum magnet, is shown. This method can serve as a pattern approach to the problem of initiating remote-handling activities in other facilities

  2. Remote-Handled Transuranic Content Codes

    International Nuclear Information System (INIS)

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document represents the development of a uniform content code system for RH-TRU waste to be transported in the 72-Bcask. It will be used to convert existing waste form numbers, content codes, and site-specific identification codes into a system that is uniform across the U.S. Department of Energy (DOE) sites.The existing waste codes at the sites can be grouped under uniform content codes without any lossof waste characterization information. The RH-TRUCON document provides an all-encompassing description for each content code and compiles this information for all DOE sites. Compliance with waste generation, processing, and certification procedures at the sites (outlined in this document foreach content code) ensures that prohibited waste forms are not present in the waste. The content code gives an overall description of the RH-TRU waste material in terms of processes and packaging, as well as the generation location. This helps to provide cradle-to-grave traceability of the waste material so that the various actions required to assess its qualification as payload for the 72-B cask can be performed. The content codes also impose restrictions and requirements on the manner in which a payload can be assembled. The RH-TRU Waste Authorized Methods for Payload Control (RH-TRAMPAC), Appendix 1.3.7 of the 72-B Cask Safety Analysis Report (SAR), describes the current governing procedures applicable for the qualification of waste as payload for the 72-B cask. The logic for this classification is presented in the 72-B Cask SAR. Together, these documents (RH-TRUCON, RH-TRAMPAC, and relevant sections of the 72-B Cask SAR) present the foundation and justification for classifying RH-TRU waste into content codes. Only content codes described in thisdocument can be considered for transport in the 72-B cask. Revisions to this document will be madeas additional waste qualifies for transport. Each content code uniquely

  3. Remote-Handled Transuranic Content Codes

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions

    2001-08-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document representsthe development of a uniform content code system for RH-TRU waste to be transported in the 72-Bcask. It will be used to convert existing waste form numbers, content codes, and site-specificidentification codes into a system that is uniform across the U.S. Department of Energy (DOE) sites.The existing waste codes at the sites can be grouped under uniform content codes without any lossof waste characterization information. The RH-TRUCON document provides an all-encompassing|description for each content code and compiles this information for all DOE sites. Compliance withwaste generation, processing, and certification procedures at the sites (outlined in this document foreach content code) ensures that prohibited waste forms are not present in the waste. The contentcode gives an overall description of the RH-TRU waste material in terms of processes and|packaging, as well as the generation location. This helps to provide cradle-to-grave traceability ofthe waste material so that the various actions required to assess its qualification as payload for the72-B cask can be performed. The content codes also impose restrictions and requirements on themanner in which a payload can be assembled.The RH-TRU Waste Authorized Methods for Payload Control (RH-TRAMPAC), Appendix 1.3.7of the 72-B Cask Safety Analysis Report (SAR), describes the current governing procedures|applicable for the qualification of waste as payload for the 72-B cask. The logic for this|classification is presented in the 72-B Cask SAR. Together, these documents (RH-TRUCON,|RH-TRAMPAC, and relevant sections of the 72-B Cask SAR) present the foundation and|justification for classifying RH-TRU waste into content codes. Only content codes described in thisdocument can be considered for transport in the 72-B cask. Revisions to this document will be madeas additional waste qualifies for transport. |Each content code uniquely

  4. ITER L 6 equatorial maintenance duct remote handling study

    International Nuclear Information System (INIS)

    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

  5. Multilayered Pipe Cutting Test for Remote Handling Maintenance

    OpenAIRE

    Haibin Chen; Jianwen Guo; Zhenzhong Sun; Xuejun Jia; Hong Tang

    2015-01-01

    Based on the requirements for remote handling maintenance (RHM) of China Spallation Neutron Source (CSNS) multilayered pipes, pipes cutting tests were performed under remote handling maintenance conditions. In this study, the results were obtained from different cutting directions and supporting intensities of pipe baseplates comparisons: When enough power was provided and the blade gripper did not slip, the cutting direction had little impact on the cutting capacity but more on the fault sur...

  6. Design of remote handling equipment for the ITER NBI

    International Nuclear Information System (INIS)

    The ITER machine has three Neutral Beam Injectors (NBIs) placed tangential to the plasma at a minimum radius of 6.25 m. During operation, neutrons produced by the D-T reactions will irradiate the NBI structure and it will become radioactive. Radiation levels will be such that all subsequent maintenance of the NBIs must be carried out remotely. The presence of tritium and possibly radioactive dust requires that precautions be taken during maintenance to prevent the escape of these contaminants beyond the prescribed boundaries. The scope of this task is both the development of remote maintenance procedures and the design of the remote handling equipment to handle the NBIs. This report describes the design of remote handling tools for the ion source and its filaments, transfer cask, maintenance time, manufacturing schedule and cost estimation. (author)

  7. A study on remote handling technology using gantry robot manipulator

    International Nuclear Information System (INIS)

    The Spent Fuel Disassembling Process Mockup(SFDPM) test facility is used for developing and testing a mechanical head end process of spent fuel, by using the PWR fuel assembly mockup. In the SFDPM test facility various equipment are installed including a rod extraction, cutting, decladding device, and a skeleton compaction device. The head end process of spent fuel assembly is used for the process of the spent fuel reuse and also, used for the interim storage process. In the SFDPM, the remote handling and control technology is developed and tested to establish the head end process. A robot manipulator is attached to the telescopic tube installed at the trolley which is movable into X and Y direction. The manipulator is used for remotely handling and transporting fuel rods, bottom nozzles, and skeletons, etc. Also, it is used for remotely cutting guide tubes in order to remove top nozzle. This paper shows the experimental results of remote handling in the SFDPM

  8. Design of remote handling equipment for the ITER NBI

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-01

    The ITER machine has three Neutral Beam Injectors (NBIs) placed tangential to the plasma at a minimum radius of 6.25 m. During operation, neutrons produced by the D-T reactions will irradiate the NBI structure and it will become radioactive. Radiation levels will be such that all subsequent maintenance of the NBIs must be carried out remotely. The presence of tritium and possibly radioactive dust requires that precautions be taken during maintenance to prevent the escape of these contaminants beyond the prescribed boundaries. The scope of this task is both the development of remote maintenance procedures and the design of the remote handling equipment to handle the NBIs. This report describes the design of remote handling tools for the ion source and its filaments, transfer cask, maintenance time, manufacturing schedule and cost estimation. (author)

  9. Evaluating ITER remote handling middleware concepts

    NARCIS (Netherlands)

    Koning, J. F.; Heemskerk, C. J. M.; Schoen, P.; Smedinga, D.; Boode, A. H.; Hamilton, D. T.

    2013-01-01

    Remote maintenance activities in ITER will be performed by a unique set of hardware systems, supported by an extensive software kit. A layer of middleware will manage and control a complex set of interconnections between teams of operators, hardware devices in various operating theatres, and databas

  10. Remote handling design for moderator-reflector maintenance in JSNS

    International Nuclear Information System (INIS)

    This report introduces the present design status of remote-handling devices for activated and used components such as moderator and reflector in a spallation neutron source of the Material and Life Science Facility (MLF) at J-PARC (Japan Proton Accelerator Research Complex). The design concept and maintenance scenario are also mentioned. A key maintenance scenario adopts that the used components should be taken out from the MLF to the other storage facility after the volume reduction of them. Almost full remote handling is available to the maintenance work except for the connection/disconnection pipes of the cooling water. Remote handling for the cooling water system is under designing and it will be prepared before being significant radiation dose by accumulation of beryllium (7Be) in future. Total six remote handling devices are used for moderator-reflector maintenance. They are also available to the proton beam window and muon target maintenance. Maintenance scenario is separated into two works. One is to replace used components to new ones during beam-stop and the other is dispose used components during beam operation. Required period of replacement work is estimated to be ∼15 days, on the other hand, the disposal work is ∼26 days after dry up work (∼30 days), respectively. Study of the maintenance scenario and the remote handling design brings about the reasonable procedures and period of the maintenance work. (author)

  11. Remote handling in nuclear fusion research

    International Nuclear Information System (INIS)

    When the Joint European Torus (JET) commences operation in 1992, the neutron flux will increase by 2 or 3 orders of magnitude activating the components of the machine to such an extent as to prohibit the access of personnel into the machine hall to carry out maintenance tasks. This paper lists operations which will have to be carried out remotely either because they are essential to the routine running of the machine or in emergencies. Remotely operated equipment which has been developed to perform these tasks is described. It is based on a system of conveyors which carry manipulators and tools to their point of operation. The principal conveyors are: a telescopic articulated mast carried on a bridge over the machine enabling tasks around and above the torus to be performed; conveyors running on rails which can reach otherwise inaccessible regions beneath the machine; an articulated arm which can position a manipulator within the torus; and a radio controlled support vehicle running on caterpillar tracks carrying a camera and tools for connecting cables to other conveyors. The main features of the control room from which the conveyors, manipulators, tools and cameras are remotely operated is also described. (UK)

  12. Remote filter handling machine for Sizewell B

    International Nuclear Information System (INIS)

    Two Filter Handling machines (FHM) have been supplied to Nuclear Electric plc for use at Sizewell B Power Station. These machines have been designed and built following ALARP principles with the functional objective being to remove radioactive filter cartridges from a filter housing and replace them with clean filter cartridges. Operation of the machine is achieved by the prompt of each distinct task via an industrial computer or the prompt of a full cycle using the automatic mode. The design of the machine features many aspects demonstrating ALARP while keeping the machine simple, robust and easy to maintain. (author)

  13. Remote-Handled Low Level Waste Disposal Project Alternatives Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2010-10-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  14. Remote handling system development of armor tile replacement for FER

    International Nuclear Information System (INIS)

    A number of armor tiles are attached to the first wall of the Fusion Experimental Reactor (FER) in order to protect the first wall against severe heat/particle loads from plasma during its operation. Although the armor tiles are made of heat-resisting materials such as graphite, they are eroded and damaged due to the loads and thus they are categorized into scheduled maintenance component. A remote handling system is required to replace a large number of tiles rapidly in the highly activated circumstance and has to be capable for adjusting a manipulator's motion taking into account a thermal deformation of the first wall and/or a positioning error of a manipulator for the remote handling system. For this purpose, a remote handling system of the armor tile replacement with a visual feedback control has been fabricated and this paper describes an experimental system and the performance test results

  15. 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.

  16. Development of nuclear fuel cycle remote handling technology

    International Nuclear Information System (INIS)

    This report presents the development of remote handling systems and remote equipment for use in the pyprocessing verification at the PRIDE (PyRoprocess Integrated inactive Demonstration facility). There are four areas conducted in this work. In first area, the prototypes of an engineering-scale high-throughput decladding voloxidizer which is capable of separating spent fuel rod-cuts into hulls and powder and collecting them separately, and an automatic equipment which is capable of collecting residual powder remaining on separated hulls were developed. In second area, a servo-manipulator system was developed to operate and maintain pyroprocess equipment located at the argon cell of the PRIDE in a remote manner. A servo-manipulator with dual arm that is mounted on the lower part of a bridge transporter will be installed on the ceiling of the in-cell and can travel the length of the ceiling. In third area, a digital mock-up and a remote handling evaluation mock-up were constructed to evaluate the pyroprocess equipments from the in-cell arrangements, remote operability and maintainability viewpoint before they are installed in the PRIDE. In last area, a base technology for remote automation of integrated pyroprocess was developed. The developed decladding voloxidizer and automatic equipment will be utilized in the development of a head-end process for pyroprocessing. In addition, the developed servo-manipulator will be used for remote operation and maintenance of the pyroprocess equipments in the PRIDE. The constructed digital mock-up and remote handling evaluation mock-up will be also used to verify and improve the pyroprocess equipments for the PRIDE application. Moreover, these remote technologies described above can be directly used in the PRIDE and applied for the KAPF (Korea Advanced Pyroprocess Facility) development

  17. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Since the amount of the spent fuel rapidly increases, the current R and D activities are focused on the technology development related with the storage and utilization of the spent fuel. In this research, to provide such a technology, the mechanical head-end process has been developed. In detail, the swing and shock-free crane and the RCGLUD(Remote Cask Grappling and Lid Unbolting Device) were developed for the safe transportation of the spent fuel assembly, the LLW drum and the transportation cask. Also, the disassembly devices required for the head-end process were developed. This process consists of an assembly downender, a rod extractor, a rod cutter, a fuel decladding device, a skeleton compactor, a force-rectifiable manipulator for the abnormal spent fuel disassembly, and the gantry type telescopic transporter, etc. To provide reliability and safety of these devices, the 3 dimensional graphic design system is developed. In this system, the mechanical devices are modelled and their operation is simulated in the virtual environment using the graphic simulation tools. So that the performance and the operational mal-function can be investigated prior to the fabrication of the devices. All the devices are tested and verified by using the fuel prototype at the mockup facility

  18. Remote handling facility and equipment used for space truss assembly

    International Nuclear Information System (INIS)

    The ACCESS truss remote handling experiments were performed at Oak Ridge National Laboratory's (ORNL's) Remote Operation and Maintenance Demonstration (ROMD) facility. The ROMD facility has been developed by the US Department of Energy's (DOE's) Consolidated Fuel Reprocessing Program to develop and demonstrate remote maintenance techniques for advanced nuclear fuel reprocessing equipment and other programs of national interest. The facility is a large-volume, high-bay area that encloses a complete, technologically advanced remote maintenance system that first began operation in FY 1982. The maintenance system consists of a full complement of teleoperated manipulators, manipulator transport systems, and overhead hoists that provide the capability of performing a large variety of remote handling tasks. This system has been used to demonstrate remote manipulation techniques for the DOE, the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan, and the US Navy in addition to the National Aeronautics and Space Administration. ACCESS truss remote assembly was performed in the ROMD facility using the Central Research Laboratory's (CRL) model M-2 servomanipulator. The model M-2 is a dual-arm, bilateral force-reflecting, master/slave servomanipulator which was jointly developed by CRL and ORNL and represents the state of the art in teleoperated manipulators commercially available in the United States today. The model M-2 servomanipulator incorporates a distributed, microprocessor-based digital control system and was the first successful implementation of an entirely digitally controlled servomanipulator. The system has been in operation since FY 1983. 3 refs., 2 figs

  19. Advanced robotic remote handling system for reactor dismantlement

    International Nuclear Information System (INIS)

    An advanced robotic remote handling system equipped with a multi-functional amphibious manipulator has been developed and used to dismantle a portion of radioactive reactor internals of an experimental boiling water reactor in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. (author)

  20. Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

    International Nuclear Information System (INIS)

    This structural acceptance criteria contains the requirements for the structural analysis and design of the Remote Handling Building (RHB) in the Tritium Extraction Facility (TEF). The purpose of this acceptance criteria is to identify the specific criteria and methods that will ensure a structurally robust building that will safely perform its intended function and comply with the applicable Department of Energy (DOE) structural requirements

  1. Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    Mertz, G.

    1999-12-16

    This structural acceptance criteria contains the requirements for the structural analysis and design of the Remote Handling Building (RHB) in the Tritium Extraction Facility (TEF). The purpose of this acceptance criteria is to identify the specific criteria and methods that will ensure a structurally robust building that will safely perform its intended function and comply with the applicable Department of Energy (DOE) structural requirements.

  2. Remote-handling challenges in fusion research and beyond

    Science.gov (United States)

    Buckingham, Rob; Loving, Antony

    2016-05-01

    Energy-producing nuclear fusion reactions taking place in tokamaks cause radiation damage and radioactivity. Remote-handling technology for repairing and replacing in-vessel components has evolved enormously over the past two decades -- and is now being deployed elsewhere too.

  3. Super-FRS Target Area Remote Handling: Scenario and Development

    Directory of Open Access Journals (Sweden)

    Luis Miguel Orona

    2013-11-01

    Full Text Available The Super-FRS, Superconducting Fragment Separator, is a unique machine that presents several challenging technical problems. One of these is regarding how to conduct maintenance in the target area where high levels of radiation will be generated and human access is forbidden. To address this problem the use of a remote maintenance system is foreseen. The objective of this paper is to develop a systems engineering (SE research and development (R&D approach suitable to develop the Super-FRS Target Area Remote Maintenance Systems (TARMS and the RH design adaptation of the components in the target area. The Super-FRS target area is described in detail in order to introduce the need for a remote maintenance system. Components in the target area are classified by adopting ITER RH maintenance classification. The general scenario of remote handling and the current target area remote maintenance system are described. Finally, the proposed systems engineering approach is presented.

  4. Preliminary design of blanket handling device and evaluation of the feasibility of eliminating the spread of radioactive contamination

    International Nuclear Information System (INIS)

    This study makes the synthesis of ongoing works concerning the remote handling device required for the handling of the internals of a fusion reactor like NET, with the double nul configuration. This work follows the analysis of the comparison of vertical versus horizontal maintenance approach

  5. The 200 l stainless steel canister - remote handling clutch assembly

    International Nuclear Information System (INIS)

    The assembly 200 l stainless steel canister with remote handling clutch is an equipment for conditioning, transport and intermediate storage of solid low- and intermediate level radioactive wastes. Loading the canister with pre-conditioned radioactive wastes is done at Post-Irradiation Examination Laboratory (LEPI) of INR Pitesti either within the transfer cell (CT) or supra-cell (SC). To this goal, lifting and handling means with which the LEPI is equipped, namely, lifting bridge and remote handling clutch are used. Conditioning of waste in view of their removal from LEPI implies their solidification in concrete and placing in stainless steel canister, the operations being effected in adequate rooms correspondingly equipped in the frame of the shop located at +8.40 m height at LEPI. Technical characteristics are: - capacity, 200 l; - external diameter, max. 600 mm; - casing height, 925 mm; casing thickness, 1.5 mm; - bottom thickness, 3 mm; - lid thickness, 3 mm. The canister cross profile of the lower and upper ends is modelled so that pilling is possible without horizontal slipping. The equipment together with remote handling clutch, engaged in a special collar of the upper part of canister, is presented

  6. Engineering the SNS RTBT/Target Interface for Remote Handling

    CERN Document Server

    Holding, M; Lang, Bonnie; Murdoch, Graeme R; Potter, Kerry G; Roseberry, Ronald T

    2005-01-01

    The SNS facility is designed for a 1.4MW 1.0GeV proton beam and the interface region of this beam with the Hg spallation target will be highly activated. This installation is located about fifteen feet below the access floor and the activity levels in the RTBT/Target interface are sufficiently high to warrant the application of Remote Handling techniques. The installed components are manufactured from radiation hard materials with serviceability beyond the lifetime of the machine, and all connections and mechanisms have been simplified to allow remote handling. The application of pneumatics to facilitate the assembly of major components and to the operation of moveable diagnostics has produced some unique design solutions.

  7. High level integration of remote handling control systems at JET

    International Nuclear Information System (INIS)

    To reduce the timescale of the JET Enhanced Performance 2 (EP2) shutdown, two multi-jointed Booms instead of one will be used for maintenance and upgrades inside the JET vessel. To fully utilize this new configuration, the control systems of the Booms have been modified at a high level to allow quick and safe interactions between them. This paper will discuss how the control systems of the Booms have been integrated to exploit the increased mechanical functionality of the Octant 1 Boom, and will demonstrate how this has improved safety, utility and efficiency for the remote handling operators during the EP2 shutdown. Other operational streamlining functions will be mentioned, as well as a look to the future of Remote Handling at JET.

  8. ITER L 7 duct remote handling equipment design report

    International Nuclear Information System (INIS)

    The operation, design and interfaces of the 'Duct Vehicle' and it's associated remote handling equipment are briefly described in this document. This equipment is being designed by Spar Aerospace Ltd. for the Divertor Test Platform as part of ITER Research and Development Project L-7. Canadian Fusion Fuels Technology Project funds this work as part of the Canadian Contribution to ITER. This document describes the equipment design status at the September 1996 design review. 23 figs

  9. Protecting worker health and safety using remote handling systems

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory (LLNL) is currently developing and installing two large-scale, remotely controlled systems for use in improving worker health and safety by minimizing exposure to hazardous and radioactive materials. The first system is a full-scale liquid feed system for use in delivering chemical reagents to LLNL's existing aqueous low-level radioactive and mixed waste treatment facility (Tank Farm). The Tank Farm facility is used to remove radioactive and toxic materials in aqueous wastes prior to discharge to the City of Livermore Water Reclamation Plant (LWRP), in accordance with established discharge limits. Installation of this new reagent feed system improves operational safety and process efficiency by eliminating the need to manually handle reagents used in the treatment processes. This was done by installing a system that can inject precisely metered amounts of various reagents into the treatment tanks and can be controlled either remotely or locally via a programmable logic controller (PLC). The second system uses a robotic manipulator to remotely handle, characterize, process, sort, and repackage hazardous wastes containing tritium. This system uses an IBM-developed gantry robot mounted within a special glove box enclosure designed to isolate tritiated wastes from system operators and minimize the potential for release of tritium to the atmosphere. Tritiated waste handling is performed remotely, using the robot in a teleoperational mode for one-of-a-kind functions and in an autonomous mode for repetitive operations. The system is compatible with an existing portable gas cleanup unit designed to capture any gas-phase tritium inadvertently released into the glove box during waste handling

  10. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

    This report describes design issues concerning remote maintenance of the ITER torus vacuum pumping system. The key issues under investigation are the valve seal exchange concept under inert gas and an alternative on-line vacuum option; flask handling support methods; flask handling/pump cell access interfacing; and valve seal inspection feasibility. The horizontal exchange of moving parts (seals/disc) for a 1.5 m regeneration isolation gate valve appears technically feasible. However, it is recommended that other commercially available valves that are lighter and narrower be examined with a view to reducing the overall size of the flask and simplifying maintenance tasks. A variant of this scheme appears feasible where the seals are replaced while the torus is under vacuum using two slit valves within the body of the main valve. This approach offers reduced cost, minimized remote handling requirements, and possibly increased plant availability. Remote handling of the flask and valve moving parts by overhead support methods is studied analytically. The forces and moments acting on the flask and resulting deflections during seal exchange operations show that a more rigid support of the flask is required than can be supplied using a crane. An alternative floor-mounted support method is proposed. Pump cell access is developed from the standpoint of the handling and transfer of a seal exchange flask as well as other pump room components. A tool for in-situ inspection of regeneration-isolation valve seats appears feasible. The concept could be developed for vacuum use as well as for in-situ repair of the seats. (21 figs.)

  11. Potential applications of advanced remote handling and maintenance technology to future waste handling facilities

    International Nuclear Information System (INIS)

    The Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) has been advancing the technology in remote handling and remote maintenance of in-cell systems planned for future US nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor are directly applicable to the in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The ORNL developments are based on the application of teleoperated force-reflecting servomanipulators controlled by an operator completely removed from the hazardous environment. These developments address the nonrepetitive nature of remote maintenance in the unstructured environments encountered in a waste handling facility. Employing technological advancements in dexterous manipulators, as well as basic design guidelines that have been developed for remotely maintained equipment and processes, can increase operation and maintenance system capabilities, thereby allowing the attainment of two Federal Waste Management System major objectives: decreasing plant personnel radiation exposure and increasing plant availability by decreasing the mean-time-to-repair in-cell maintenance and process equipment

  12. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

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

  13. Remote-handled/special case TRU waste characterization summary

    International Nuclear Information System (INIS)

    TRU wastes are those (other than high level waste) contaminated with specified quantities of certain alpha-emitting radionuclides of long half-life and high specific radiotoxicity. TRU waste is defined as 226Ra isotopic sources and those other materials that, without regard to source or form, are contaminated with transuranic elements with half-lives greater than 20 years, and have TRU alpha contamination greater than 100 nCi/g. RH TRU waste has high beta and gamma radiation levels, up to 30,000 R/hr, and thermal output may be a few hundred watts per container. The radiation levels in most of this remotely handled (RH) TRU waste, however, are below 100 R/hr. Remote-handled wastes are stored at Los Alamos, Hanford, Oak Ridge, and the Idaho National Engineering Laboratory. This report presents a site by site discussion of RH waste handling, placement, and container data. This is followed by a series of data tables that were compiled in the TRU Waste Systems Office. These tables are a compendium of data that are the most up to date and accurate data available today. 10 tables

  14. Remote waste handling and feed preparation for Mixed Waste Management

    International Nuclear Information System (INIS)

    The Mixed Waste Management Facility (MWMF) at the Lawrence Livermore National Laboratory (LLNL) will serve as a national testbed to demonstrate mature mixed waste handling and treatment technologies in a complete front-end to back-end --facility (1). Remote operations, modular processing units and telerobotics for initial waste characterization, sorting and feed preparation have been demonstrated at the bench scale and have been selected for demonstration in MWMF. The goal of the Feed Preparation design team was to design and deploy a robust system that meets the initial waste preparation flexibility and productivity needs while providing a smooth upgrade path to incorporate technology advances as they occur. The selection of telerobotics for remote handling in MWMF was made based on a number of factors -- personnel protection, waste generation, maturity, cost, flexibility and extendibility. Modular processing units were selected to enable processing flexibility and facilitate reconfiguration as new treatment processes or waste streams are brought on line for demonstration. Modularity will be achieved through standard interfaces for mechanical attachment as well as process utilities, feeds and effluents. This will facilitate reconfiguration of contaminated systems without drilling, cutting or welding of contaminated materials and with a minimum of operator contact. Modular interfaces also provide a standard connection and disconnection method that can be engineered to allow convenient remote operation

  15. The ITER EC H and CD upper launcher: Analysis of remote handling compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Ronden, D.M.S., E-mail: d.m.s.ronden@rijnhuizen.nl [FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, partner in the Trilateral Euregio Cluster and ITER-NL, P.O. 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, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Chavan, R. [CRPP, EURATOM-Confederation Suisse, EPFL, CH-1015 Lausanne (Switzerland); Elzendoorn, B.S.Q. [FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, partner in the Trilateral Euregio Cluster and ITER-NL, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Goodman, T. [CRPP, EURATOM-Confederation Suisse, EPFL, CH-1015 Lausanne (Switzerland); Heemskerk, C.J.M. [Heemskerk Innovative Technology, Merelhof 2, 2172 HZ, Sassenheim (Netherlands); Henderson, M.A. [ITER-IO, Cadarache 13108 Saint Paul Lez Durance (France); Koning, J.F. [FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, partner in the Trilateral Euregio Cluster and ITER-NL, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Saibene, G. [FUSION FOR ENERGY, Joint Undertaking, 08019 Barcelona (Spain); Spaeh, P.; Strauss, D. [Karlsruhe Institute of Technology, Association KIT-EURATOM, Institute for Materials Research I, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2011-10-15

    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.

  16. Remote automated material handling of radioactive waste containers

    International Nuclear Information System (INIS)

    To enhance personnel safety, improve productivity, and reduce costs, the design team incorporated a remote, automated stacker/retriever, automatic inspection, and automated guidance vehicle for material handling at the Enhanced Radioactive and Mixed Waste Storage Facility - Phase V (Phase V Storage Facility) on the Hanford Site in south-central Washington State. The Phase V Storage Facility, scheduled to begin operation in mid-1997, is the first low-cost facility of its kind to use this technology for handling drums. Since 1970, the Hanford Site's suspect transuranic (TRU) wastes and, more recently, mixed wastes (both low-level and TRU) have been accumulating in storage awaiting treatment and disposal. Currently, the Hanford Site is only capable of onsite disposal of radioactive low-level waste (LLW). Nonradioactive hazardous wastes must be shipped off site for treatment. The Waste Receiving and Processing (WRAP) facilities will provide the primary treatment capability for solid-waste storage at the Hanford Site. The Phase V Storage Facility, which accommodates 27,000 drum equivalents of contact-handled waste, will provide the following critical functions for the efficient operation of the WRAP facilities: (1) Shipping/Receiving; (2) Head Space Gas Sampling; (3) Inventory Control; (4) Storage; (5) Automated/Manual Material Handling

  17. Three-dimensional television system for remote handling

    International Nuclear Information System (INIS)

    The paper refers to work previously described on the development of 3-D Television Systems. 3-D TV had been developed with a view to proving whether it was a useful remote handling tool which would be easy to use and comfortable to view. The paper summarizes the work of evaluation trials at UK facilities and reviews the developments which have subsequently taken place. 3-D TV systems have been found to give improved performance in terms of speed and accuracy of operations and to reduce the number of camera views required. (author)

  18. Remote handling and robotics at the BNFL Sellafield reprocessing plant

    International Nuclear Information System (INIS)

    As a direct result of its interest in the use of robotics within active plants, British Nuclear Fuels Ltd. (BNFL) has adopted a positive attitude toward both national and European initiatives in this area. During the early operation of the Sellafield reprocessing plant, the process vessels and cell voids were monitored using simple pole and camera combinations. In 1985, BNFL embarked on the provision of a series of machines intended to satisfy the advancing needs for inspection while increasing the level of expertise within the company in this important area. DIMAN 1, DIMAN 2, RODMAN, REPMAN, and RAFFMAN remote handling and robotic machines are described

  19. Nuclear robotics and remote handling at Harwell Laboratory

    International Nuclear Information System (INIS)

    After reviewing robotics technology and its possible application in nuclear remote handling systems of the future, six main research topics were identified where particular effort should be made. The Harwell Nuclear Robotics Programme is currently establishing sets of demonstration hardware which will allow generic research to be carried out on telerobotics, systems integration, the man machine interface, communications, servo systems and radiation tolerance. The objectives of the demonstrators are to allow validation of the techniques required for successful active facility applications such as decommissioning, decontamination, refurbishment, maintenance and repair, and to act as training aids to encourage plant designers and operators to adopt developments in new technology. (author)

  20. Development of nuclear fuel cycle remote handling technology

    International Nuclear Information System (INIS)

    This report presents the development of remote handling systems and remote equipment for use in the pyprocessing verification at the PRIDE (PyRoprocess Integrated inactive Demonstration facility). There are three areas conducted in this work. In first area, developed were the prototypes of an engineering-scale high-throughput decladding voloxidizer which is capable of separating spent fuel rod-cuts into hulls and powder and collecting them separately and an automatic equipment which is capable of collecting residual powder remaining on separated hulls. In second area, a servo-manipulator prototype was developed to operate and maintain pyroprocess equipment located at the argon cell of the PRIDE in a remote manner. A servo-manipulator with dual arm that is mounted on the lower part of a bridge transporter will be installed on the ceiling of the in-cell and can travel the length of the ceiling. In last area, a simulator was developed to simulate and evaluate the design developments of the pyroprocess equipment from the in-cell arrangements, remote operability and maintainability viewpoint in a virtual process environment in advance before they are constructed. The developed decladding voloxidizer and automatic equipment will be utilized in the development of a head-end process for pyroprocessing. In addition, the developed servo-manipulator will be installed in the PRIDE and used for remote operation and maintenance of the pyroprocess equipment. The developed simulator will be also used to verify and improve the design of the pyroprocess equipment for the PRIDE application. Moreover, these remote technologies described above can be directly used in the PRIDE and applied for the ESPF (Engineering Scale Pyroprocess Facility) and KAPF (Korea Advanced Pyroprocess Facility) development

  1. A dynamic simulation for the study of remote handling operations

    International Nuclear Information System (INIS)

    The inspection, maintenance and repair of fusion machines will require the extensive use of remote handling equipment to minimise the human exposure to the high radiation environment. A high fidelity simulation can be a very valuable tool to assist in the manipulator design, operations, trajectory planning, parameter optimisation and system verification. ASAD, a non-real time simulation program developed by Spar Aerospace, through funding from the National Research Council of Canada, is used to simulate the remote manipulator on the US Space Shuttle program. This simulation program is now being adapted to simulate a wide range of terrestrial manipulator configurations, most notably the In-Vessel Handling Unit (IVHU) for the Next European Torus (NET) program. This terrestrial version ASAD-T, is capable of simulating a manipulator consisting of up to 7 joints, and includes models of the flexible dynamics of the drive system and the structure, as well as the details of the control algorithms at the manipulator level and at the joint level. A modular approach has been used such that a variety of manipulators and control systems can be simulated by module substitution. Extensive use of parametric data to describe the manipulator and control system permits the modelling of a wide variety of manipulator configurations. The simulation, as configured for the Space Shuttle Remote Manipulator System, has undergone an extensive validation process against other simulations, test data and actual flight data. This paper describes the capabilities and underlying assumptions of the simulation program ASAD-T, along with a brief description of the simulation development of an in-vessel manipulator for fusion reactors now in progress. (author). 4 refs, 8 figs

  2. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

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

  3. Survey on Remote Handling Logistics for Super-FRS

    Directory of Open Access Journals (Sweden)

    Faraz Amjad

    2013-10-01

    Full Text Available Remote Handling (RH systems are now frequently used to conduct inspections and maintenance in hazardous environments. New particle accelerator facilities present unique logistic challenges due to high radiation levels, a hazardous environment and heavy loads. The Facility for Antiproton and Ion Research (FAIR will deliver a beam of all ions up to uranium with intensities up to 1012 238U ions/s, which will cause high levels of radiation during operation so human access is limited. This paper contains a survey on RH logistics for existing High Intensity Beam (HIB facilities to determine state of the art RH practices and to draw a conclusion based on the analysis. The second part of this paper presents a detailed study of beam losses, the radiation environment, RH logistic challenges and some proposed solutions for Super- FRS. This paper will also suggest a Systems Engineering (SE approach for developing Super-FRS RH logistics.

  4. Remote Inspection, Measurement and Handling for Maintenance and Operation at CERN

    OpenAIRE

    Keith Kershaw; Bruno Feral; Jean-Louis Grenard; Thierry Feniet; Sven De Man; Cathelijne Hazelaar-Bal; Caterina Bertone; Ruehl Ingo

    2013-01-01

    Remote inspection, measurement and handling techniques are under development for use in CERN’s particle accelerator facilities to reduce radiation exposure of personnel and reduce facility downtime when scheduled maintenance or breakdown repairs are needed. This paper gives an explanation of the potential benefits to CERN of remote inspection, measurement and handling along with a brief history of remote handling work at CERN. Recent projects are then described, covering the development work ...

  5. Reliability of robotics: An overview with identification of specific aspects related to remote handling in fusion machines

    International Nuclear Information System (INIS)

    Nowadays robots are indispensable in a major part of industry branches. Robots have the ability to perform precision demanding tasks at high speed and uninterruptedly over a long period of time, and to operate in environments which are inaccessible or hostile for human beings. In general, robots can even be considered as being more reliable than humans, and robots improve continuously thanks to ever more sophisticated technology. Therefore, to a certain extent, reliability and availability of robots can be taken for granted, and efforts of robot designers concentrate now more and more on safety, i.e. the risk robots may constitute for persons. Safety, as e.g. defined for robots working in production sites, differs slightly from safety related to remote handling devices in fusion machines, which operate in a contaminated environment where no person is supposed to be. The term safety should therefore not only consider the potentiality of human loss, but must be extended to severe damage of machinery and equipment. This paper describes a safety analysis methodology for robotized mechanical systems. This method has been applied on a case study, a 1/3 scale mock-up of a handling device for the blankets of the fusion reactor segments (ROBERTINO). In particular were analyzed two different mechanical concepts of gripping the blankets

  6. Progress in standardization for ITER Remote Handling control system

    International Nuclear Information System (INIS)

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

  7. Remote-Handled Low-Level Waste (RHLLW) Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2010-10-01

    The Remote-Handled Low-Level Waste Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of fiscal year 2015). Development of a new onsite disposal facility, the highest ranked alternative, will provide necessary remote handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability.

  8. Remote Inspection, Measurement and Handling for Maintenance and Operation at CERN

    Directory of Open Access Journals (Sweden)

    Keith Kershaw

    2013-11-01

    Full Text Available Remote inspection, measurement and handling techniques are under development for use in CERN’s particle accelerator facilities to reduce radiation exposure of personnel and reduce facility downtime when scheduled maintenance or breakdown repairs are needed. This paper gives an explanation of the potential benefits to CERN of remote inspection, measurement and handling along with a brief history of remote handling work at CERN. Recent projects are then described, covering the development work and operations forming part of CERN’s recent remote inspection, measurement and handling activities.

  9. The National Remote-Handled Transuranic Waste Program

    International Nuclear Information System (INIS)

    Since the Waste Isolation Pilot Plant (WIPP) opened for nonmixed contact-handled (CH) transuranic (TRU) waste receipt on March 26, 1999, planning for the receipt of remote-handled (RH) TRU waste has been revitalized. WIPP is scheduled to begin receiving RH TRU waste in January 2002. Before the first receipt of RH waste at WIPP, the disposal of RH TRU waste must be coordinated with the waste management plans at the TRU waste sites, the packaging and transportation system must be in place and operational, and the WIPP RH must be operational. Successful implementation of the National RH TRU Waste Program will require that a commitment to sustained funding and full support of the entire DOE TRU waste complex be maintained. Implementation will also require effective application of capital assets, human resources, and a continued effort to identify and resolve programmatic and technical issues. Involvement of regulatory agencies, stakeholders, and affected tribal nations in decision-making processes will have a positive impact on successful implementation

  10. The decontamination of IFD containment using remote handling techniques

    International Nuclear Information System (INIS)

    This paper discusses the method by which a specific decontamination requirement for the Irradiated Fuel Dismantling (IFD) Cell No. 2 at Torness Power Station was addressed. The design brief required conventional manually operated equipment to be used and the design process used a computer modelling tool to develop the equipment design and to analyse its operation. The paper reviews the background to the decontamination requirement in the Upper Containment Box (UCB) of the cell. All handling activities and tie bar cutting operations are undertaken in this area resulting in the UCB floor becoming a focal point for contamination. The original decontamination equipment proved inadequate owing to changes in the UCB and, consequently, manual decontamination became necessary. Principal considerations for the introduction of equipment to the containment were the availability of suitable access points, the limited operating space and the normal operational requirements of the cell equipment itself. The design involved using a Master Slave Manipulator (MSM) as a remote handling tool which controlled and directed a support arm carrying a vacuum nozzle. (author)

  11. ITER Equatorial Port plug engineering: Design and remote handling activities supported by Virtual Reality tools

    International Nuclear Information System (INIS)

    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.

  12. Defense Remote Handled Transuranic Waste Cost/Schedule Optimization Study

    International Nuclear Information System (INIS)

    The purpose of this study is to provide the DOE information with which it can establish the most efficient program for the long management and disposal, in the Waste Isolation Pilot Plant (WIPP), of remote handled (RH) transuranic (TRU) waste. To fulfill this purpose, a comprehensive review of waste characteristics, existing and projected waste inventories, processing and transportation options, and WIPP requirements was made. Cost differences between waste management alternatives were analyzed and compared to an established baseline. The result of this study is an information package that DOE can use as the basis for policy decisions. As part of this study, a comprehensive list of alternatives for each element of the baseline was developed and reviewed with the sites. The principle conclusions of the study follow. A single processing facility for RH TRU waste is both necessary and sufficient. The RH TRU processing facility should be located at Oak Ridge National Laboratory (ORNL). Shielding of RH TRU to contact handled levels is not an economic alternative in general, but is an acceptable alternative for specific waste streams. Compaction is only cost effective at the ORNL processing facility, with a possible exception at Hanford for small compaction of paint cans of newly generated glovebox waste. It is more cost effective to ship certified waste to WIPP in 55-gal drums than in canisters, assuming a suitable drum cask becomes available. Some waste forms cannot be packaged in drums, a canister/shielded cask capability is also required. To achieve the desired disposal rate, the ORNL processing facility must be operational by 1996. Implementing the conclusions of this study can save approximately $110 million, compared to the baseline, in facility, transportation, and interim storage costs through the year 2013. 10 figs., 28 tabs

  13. Simulating and visualizing deflections of a remote handling mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Saarinen, Hannu, E-mail: hannu.saarinen@vtt.fi [VTT, Technical Research Centre of Finland, Tekniikankatu 1, 33720 Tampere (Finland); Hämäläinen, Vesa; Karjalainen, Jaakko; Määttä, Timo; Siuko, Mikko [VTT, Technical Research Centre of Finland, Tekniikankatu 1, 33720 Tampere (Finland); Esqué, Salvador [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Hamilton, David [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► An infinitesimal transformation represents elastic deflections. ► Equivalent spring factor is used to combine several deformations. ► Initial VR model accuracy improved from 80 to 5 mm. ► The deflection model is capable of adapting to changes in load at the end-effector. ► The algorithms and approach described are generic and can be adopted for other mechanisms. -- Abstract: Continuing ITER divertor second cassette (SC) remote handling (RH) test campaign has been carried out at divertor test platform (DTP2) in Finland. One of the goals has been to develop and implement efficient algorithms and software tools for simulating and visualizing for the operator the non-instrumented deflections of the RH mechanisms under loading conditions. Based on assumptions of the classical beam theory, the presented solution suggests utilization of an infinitesimal transformation to represent elastic deflections in a mechanical structure. Both structural analysis and measurements of the real structure are utilised during the process. The solution suggests one possible implementation strategy of a software component called structural simulator (SS), which is a software component of the remote handling control system (RHCS) architectural model specified by ITER organisation. Utilisation of the proposed SS necessitates modification of the initial virtual reality (VR) model of RH equipment to a format, which can visually represent the structural deflections. In practise this means adding virtual joints into the model. This will improve the accuracy of the VR visualization and will ensure that the virtual representation of the RH equipment closely aligns with the actual RH equipment. Cassette multifunctional mover (CMM) and second cassette end effector (SCEE) carrying SC were selected to be the initial target system for developing the approach. Demonstrations proved that the approach used can give high levels of accuracy even in complex structures such as the CMM

  14. Survey of technology for decommissioning of nuclear fuel cycle facilities. 8. Remote handling and cutting techniques

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Ryuichiro; Ishijima, Noboru [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-03-01

    In nuclear fuel cycle facility decommissioning and refurbishment, the remote handling techniques such as dismantling, waste handling and decontamination are needed to reduce personnel radiation exposure. The survey research for the status of R and D activities on remote handling tools suitable for nuclear facilities in the world and domestic existing commercial cutting tools applicable to decommissioning of the facilities was conducted. In addition, the drive mechanism, sensing element and control system applicable to the remote handling devices were also surveyed. This report presents brief surveyed summaries. (H. Itami)

  15. Radioactive package container system for remote handling of low-level radioactive waste

    International Nuclear Information System (INIS)

    Remotely operated handling systems are employed for safe processing and transfer of low level radioactive wastes at nuclear generating plants. These systems minimize or preclude personnel radiation exposure while expediting waste handling operations. A remotely operated waste handling and transfer system containing several unique features has been designed, fabricated and tested for installation oat Arizona Public Service's, Palo Verde Nuclear Generating Station. The system incorporates modular subcomponents such as a waste processing shield, bottom and top loading shielded cask, and remote grappling equipment, making it adaptable to multi-task waste handling operations. The system has been designed to be operationally flexible, and contributes significantly to reducing waste processing personnel exposure

  16. On-site transfer system for remote handling of low-level radioactive waste

    International Nuclear Information System (INIS)

    Remotely operated handling systems are employed for safe processing and transfer of low level radioactive wastes at nuclear generating plants. These systems minimize or preclude personnel radiation exposure while expediting waste handling operations. A remotely operated waste handling and transfer system containing several unique features has been designed, fabricated and installed at Southern California Edison's, San Onofre Nuclear Generating Station. The system incorporates modular subcomponents such as a waste processing shield, bottom and top loading shielded cask, transportation system and remote grappling equipment, making it adaptable to multi-task waste handling operations. The system has proven to be operationally flexible, and has contributed significantly to reducing waste processing personnel exposure

  17. Remote material handling in the Plutonium Immobilization Project. Revision 1

    International Nuclear Information System (INIS)

    With the downsizing of the US and Russian nuclear stockpiles, large quantities of weapons-usable plutonium in the US are being declared excess and will be disposed of by the Department of Energy Fissile Materials Disposition Program. To implement this program, DOE has selected the Savannah River Site (SRS) for the construction and operation of three new facilities: pit disassembly and conversion; mixed oxide fuel fabrication; and plutonium immobilization. The Plutonium Immobilization Project (PIP) will immobilize a portion of the excess plutonium in a hybrid ceramic and glass form containing high level waste for eventual disposal in a geologic repository. The PIP is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. Processing technology for the PIP is being developed jointly by the Lawrence Livermore National Laboratory and Westinghouse Savannah River Company. This paper will discuss development of the automated unpacking and sorting operations in the conversion area, and the automated puck and tray handling operations in the first stage immobilization area. Due to the high radiation levels and toxicity of the materials to be disposed of, the PIP will utilize automated equipment in a contained (glovebox) facility. Most operations involving plutonium-bearing materials will be performed remotely, separating personnel from the radiation source. Source term materials will be removed from the operations during maintenance. Maintenance will then be performed hands on within the containment using glove ports

  18. Remote handling devices for radioactive materials - Part 4: Power manipulators

    International Nuclear Information System (INIS)

    This part of ISO 17874 deals with power manipulators used for nuclear applications. These manipulators consist mainly of multipurpose remote handling devices. These devices replace hands and arms and even light hoists, depending on the model used, in areas inaccessible to personnel (mostly behind shielding walls). Power manipulators were originally developed for hot cells designed for research and development in fuel elements for nuclear power reactors. They are now also in widespread use in other nuclear installations, such as plants for reprocessing of fuel elements, waste treatment stations, and decommissioning of nuclear facilities. Alternative manipulators used in these fields and resulting in a wide variety of different designs are considered to be skill in an emergent phase or applied uniquely in special circumstances and are not addressed further in this current edition of this standard. Power manipulators are sometimes modified or especially designed for non-nuclear applications. This part of ISO 17874 does not address the special requirements of any of these applications. Although designers may not be taken advantage of standardized features and components from the nuclear sector to achieve efficient and cost-effective designs for other purposes where appropriate. This part of ISO 17874 is intended to provide assistance to designers of nuclear process and research plants, as well as manufacturers, users and licensing authorities

  19. Remote controlled stud bolt handling device for reactor pressure vessel

    International Nuclear Information System (INIS)

    In nuclear power stations, at the time of regular inspection, the works of opening and fixing the upper covers of reactor pressure vessels are carried out for inspecting the inside of reactor pressure vessels and exchanging fuel rods. These upper covers are fastened with many stud bolts, therefore, the works of opening and fixing require a large amount of labor, and are done under the restricted condition of wearing protective clothings and masks. Babcock Hitachi K.K. has completed the development of a remotely controlled automatic bolt tightenig device for this purpose, therefore, its outline is reported. The conventional method of these works and the problems in it are described. The design of the new device aimed at the parallel execution of cleaning screw threads, loosening and tightening nuts, and taking off and putting on nuts and washers, thus contributing to the shortening of regular inspection period, the reduction of the radiation exposure of workers, and the decrease of the number of workers. The function, reliability and endurance of the new device were confirmed by the verifying test using a device made for trial. The device is composed of a stand, a rail and four stations each with a cleaning unit, a stud tensioner and a nut handling unit. (K.I.)

  20. B cell remote-handled waste shipment cask alternatives study

    International Nuclear Information System (INIS)

    The decommissioning of the 324 Facility B Cell includes the onsite transport of grouted remote-handled radioactive waste from the 324 Facility to the 200 Areas for disposal. The grouted waste has been transported in the leased ATG Nuclear Services 3-82B Radioactive Waste Shipping Cask (3-82B cask). Because the 3-82B cask is a U.S. Nuclear Regulatory Commission (NRC)-certified Type B shipping cask, the lease cost is high, and the cask operations in the onsite environment may not be optimal. An alternatives study has been performed to develop cost and schedule information on alternative waste transportation systems to assist in determining which system should be used in the future. Five alternatives were identified for evaluation. These included continued lease of the 3-82B cask, fabrication of a new 3-82B cask, development and fabrication of an onsite cask, modification of the existing U.S. Department of Energy-owned cask (OH-142), and the lease of a different commercially available cask. Each alternative was compared to acceptance criteria for use in the B Cell as an initial screening. Only continued leasing of the 3-82B cask, fabrication of a new 3-82B cask, and the development and fabrication of an onsite cask were found to meet all of the B Cell acceptance criteria

  1. Development of a Remote Handling System in an Integrated Pyroprocessing Facility

    OpenAIRE

    Hyo Jik Lee; Jong Kwang Lee; Byung Suk Park; Kiho Kim; Won Il Ko; Il Je Cho

    2013-01-01

    Over the course of a decade-long research programme, the Korea Atomic Energy Research Institute (KAERI) has developed several remote handling systems for use in pyroprocessing research facilities. These systems are now used successfully for the operation and maintenance of processing equipment. The most recent remote handling system is the bridge-transported dual arm servo-manipulator system (BDSM), which is used for remote operation at the world’s largest pyroprocess integrated inactive demo...

  2. Preliminary Hazard Analysis for the Remote-Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; Mike Lehto

    2010-05-01

    The need for remote handled low level waste (LLW) disposal capability has been identified. A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal capability for remote-handled LLW that is generated as part of the nuclear mission of the Idaho National Laboratory and from spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This document supports the conceptual design for the proposed remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization and by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW.

  3. Preliminary Hazard Analysis for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; Mike Lehto

    2010-10-01

    The need for remote handled low level waste (LLW) disposal capability has been identified. A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal capability for remote-handled LLW that is generated as part of the nuclear mission of the Idaho National Laboratory and from spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This document supports the conceptual design for the proposed remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization and by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW.

  4. Project Execution Plan for the Remote Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Danny Anderson

    2014-07-01

    As part of ongoing cleanup activities at the Idaho National Laboratory (INL), closure of the Radioactive Waste Management Complex (RWMC) is proceeding under the Comprehensive Environmental Response, Compensation, and Liability Act (42 USC 9601 et seq. 1980). INL-generated radioactive waste has been disposed of at RWMC since 1952. The Subsurface Disposal Area (SDA) at RWMC accepted the bulk of INL’s contact and remote-handled low-level waste (LLW) for disposal. Disposal of contact-handled LLW and remote-handled LLW ion-exchange resins from the Advanced Test Reactor in the open pit of the SDA ceased September 30, 2008. Disposal of remote-handled LLW in concrete disposal vaults at RWMC will continue until the facility is full or until it must be closed in preparation for final remediation of the SDA (approximately at the end of fiscal year FY 2017). The continuing nuclear mission of INL, associated ongoing and planned operations, and Naval spent fuel activities at the Naval Reactors Facility (NRF) require continued capability to appropriately dispose of contact and remote handled LLW. A programmatic analysis of disposal alternatives for contact and remote-handled LLW generated at INL was conducted by the INL contractor in Fiscal Year 2006; subsequent evaluations were completed in Fiscal Year 2007. The result of these analyses was a recommendation to the Department of Energy (DOE) that all contact-handled LLW generated after September 30, 2008, be disposed offsite, and that DOE proceed with a capital project to establish replacement remote-handled LLW disposal capability. An analysis of the alternatives for providing replacement remote-handled LLW disposal capability has been performed to support Critical Decision-1. The highest ranked alternative to provide this required capability has been determined to be the development of a new onsite remote-handled LLW disposal facility to replace the existing remote-handled LLW disposal vaults at the SDA. Several offsite DOE

  5. Nondestructive assay and nondestructive examination of remote-handled transuranic waste at the ORNL waste handling and packaging plant

    International Nuclear Information System (INIS)

    The purpose of this investigation is to examine the use of an electron linear accelerator (LINAC) in the performance of nondestructive assay (NDA) and nondestructive examination (NDE) measurements of remote-handled transuranic wastes. The system will be used to perform waste characterization and certification activities at the Oak Ridge National Laboratory's proposed Waste Handling and Packaging Plant. The NDA and NDE technologies which were developed for contact-handled wastes are inadequate to perform such measurements on high gamma and neutron dose-rate wastes. A single LINAC will provide the interrogating fluxes required for both NDA and NDE measurements of the wastes. 11 refs., 6 figs

  6. Nondestructive assay and nondestructive examination of remote-handled transuranic waste at the ORNL waste handling and packaging plant

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, F.J.; Caldwell, J.T. (Oak Ridge National Lab., TN (USA); Pajarito Scientific Corp. (USA))

    1989-01-01

    The purpose of this investigation is to examine the use of an electron linear accelerator (LINAC) in the performance of nondestructive assay (NDA) and nondestructive examination (NDE) measurements of remote-handled transuranic wastes. The system will be used to perform waste characterization and certification activities at the Oak Ridge National Laboratory's proposed Waste Handling and Packaging Plant. The NDA and NDE technologies which were developed for contact-handled wastes are inadequate to perform such measurements on high gamma and neutron dose-rate wastes. A single LINAC will provide the interrogating fluxes required for both NDA and NDE measurements of the wastes. 11 refs., 6 figs.

  7. Conceptual Safety Design Report for the Remote Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Christensen

    2010-05-01

    A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal for remote-handled LLW from the Idaho National Laboratory and for spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This conceptual safety design report supports the design of a proposed onsite remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization, by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW, by evaluating consequences of postulated accidents, and by discussing the need for safety features that will become part of the facility design.

  8. Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Solack; Carol Mason

    2012-03-01

    A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

  9. Advanced remote handling for future applications: The advanced integrated maintenance system

    International Nuclear Information System (INIS)

    The Consolidated Fuel Reprocessing Program at Oak Ridge National Laboratory has been developing advanced techniques for remote maintenance of future US fuel reprocessing plants. The developed technology has a wide spectrum of application for other hazardous environments. These efforts are based on the application of teleoperated, force-reflecting servomanipulators for dexterous remote handling with television viewing for large-volume hazardous applications. These developments fully address the nonrepetitive nature of remote maintenance in the unstructured environments encountered in fuel reprocessing. This paper covers the primary emphasis in the present program; the design, fabrication, installation, and operation of a prototype remote handling system for reprocessing applications, the Advanced Integrated Maintenance System

  10. Applying Remote Handling Attributes to the ITER Neutral Beam Cell Monorail Crane

    CERN Document Server

    Crofts, O; Raimbach, J; Tesini, A; Choi, C-H; Damiani, C; Van Uffelen, M

    2013-01-01

    The maintenance requirements for the equipment in the ITER Neutral Beam Cell requires components to be lifted and transported within the cell by remote means. To meet this requirement, the provision of an overhead crane with remote handling capabilities has been initiated. The layout of the cell has driven the design to consist of a monorail crane that travels on a branched monorail track attached to the cell ceiling. This paper describes the principle design constraints and how the remote handling attributes were applied to the concept design of the monorail crane, concentrating on areas where novel design solutions have been required and on the remote recovery requirements and solutions.

  11. Applying remote handling attributes to the ITER neutral beam cell monorail crane

    International Nuclear Information System (INIS)

    The maintenance requirements for the equipment in the ITER neutral beam cell require components to be lifted and transported within the cell by remote means. To meet this requirement, the provision of an overhead crane with remote handling capabilities has been initiated. The layout of the cell has driven the design to consist of a monorail crane that travels on a branched monorail track attached to the cell ceiling. This paper describes the principle design constraints and how the remote handling attributes were applied to the concept design of the monorail crane, concentrating on areas where novel design solutions have been required and on the remote recovery requirements and solutions

  12. Development of monitoring-control methods for heavy remote handling operations in an irradiated environment

    International Nuclear Information System (INIS)

    Heavy remote handling equipment units have benefited from the progress made in robotics, but with certain specific constraints linked to the environment in which they are required to operate. Notably, these constraints impose the exclusive use of electrical techniques

  13. Irradiation tests of critical components for remote handling in gamma radiation environment

    International Nuclear Information System (INIS)

    Since the fusion power core of a D-T fusion reactor will be highly activated once it starts operation, personnel access will be prohibited so that assembly and maintenance of the components in the reactor core will have to be totally conducted by remote handling technology. Fusion experimental reactors such as ITER require unprecedented remote handling equipments which are tolerable under gamma radiation of more than 106 R/h. For this purpose, the Japan Atomic Energy Research Institute (JAERI) has been developing radiation hard components for remote handling purpose and a number of key components have been tested over 109 rad at a radiation dose rate of around 106 R/h, using Gamma Ray Radiation Test Facility in JAERI-Takasaki Establishment. This report summarizes the irradiation test results and the latest status of AC servo motor, potentiometer, optical elements, lubricant, sensors and cables, which are key elements of the remote handling system. (author)

  14. Remote-Handled Low-Level Waste Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2011-04-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility, the highest ranked alternative, will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  15. Remote-Handled Low-Level Waste Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2011-01-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility, the highest ranked alternative, will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  16. Remote-Handled Low-Level Waste Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2012-04-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  17. Remote-Handled Low-Level Waste Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    Austad, S. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Guillen, L. E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); McKnight, C. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ferguson, D. S. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  18. Remote-Handled Low-Level Waste Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2012-06-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  19. Remote-Handled Low-Level Waste Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2014-06-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  20. Remote system development for demonstrating compliance with remote-handled waste disposal criteria

    International Nuclear Information System (INIS)

    On March 25, 1999, the Waste Isolation Pilot Plant (WIPP) received its first shipment of contact-handled (CH) transuranic (TRU) waste from the Los Alamos National Laboratory for disposal within a geologic formation of bedded salt approximately 2150 ft underground. The WIPP is owned and operated by the US Department of Energy (DOE) and is located in a remote location at the southeast corner of New Mexico. Since that opening date, additional shipments of CH-TRU waste have commenced from other generator/storage sites including the Idaho National Engineering and Environmental Laboratory (April 28, 1999) and Rocky Flats Environmental Technology Site (June 16, 1999). These initial shipments contained TRU waste that was determined by the generator/storage sites to have a surface dose rate not greater than 2 mSv/h and to be nonmixed (i.e., contained no hazardous constituents regulated under the Resource Conservation and Recovery Act). On November 26, 1999, the WIPP Hazardous Waste Facility Permit was issued by the New Mexico Environment Department (NMED). Although this permit greatly expanded the waste envelope by allowing the disposal of mixed waste at the WIPP, it effectively prohibited the disposal of remote-handled (RH) TRU waste (i.e., waste having a surface dose rate >2 mSv/h but le10 Sv/h). As a consequence, the WIPP has embarked on a path that will facilitate the disposal of RH-TRU waste by the year 2003. One of the milestones along this path is the establishment of the WIPP waste acceptance criteria for the disposal of RH-TRU waste (mixed and nonmixed)

  1. Conceptualization of a generic remote handling system for tokamak maintenance applications

    International Nuclear Information System (INIS)

    Remote handling will have an important role in the operation of the fusion machines. When operation begins, it will be impossible to make changes, conduct inspections, or repair any of the Tokamak components in the activated areas other than by remote handling. Very reliable and robust remote handling techniques will be necessary to manipulate and exchange components. With high temperatures and radiation levels and the huge work space, much of the inspection and maintenance tasks would be carried out by articulated manipulators. The Tokamak remote handling (RH) maintenance system is a key component in Tokamak 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, nuclearized design. This paper presents a concept for a generic remote handling system which can cater to various repairing, installation and maintenance requirements of a tokamak device. The system is divided into several modules like Deployer, Multi-Purpose Manipulator, Task Module, Transfer and Service casks. Various RH end effectors and tools can be mounted on the manipulator to perform maintenance tasks such as cleaning of the In-Vessel components, heavy material handling, In-Vessel viewing and Inspection etc. The design and analysis methodology based on the kinematic parameters, Servo Joint mechanisms, and Gear based mechanisms is presented. (author)

  2. Development of in-vessel remote maintenance system

    International Nuclear Information System (INIS)

    In the ITER system, such in-vessel components as blanket and divertor are treated as schedule maintenance components, and the remote handling equipment/tools required for in-vessel maintenance must be capable of handling a heavy payload with high precision, even under intense gamma ray radiation. To facilitate remote maintainability, the blanket and divertor are structurally segmented into modules and cassettes. In addition, a rail-mounted vehicle-type remote handling system and a cassette-type remote handling system have been developed for blanket and divertor maintenance, respectively. Critical technology relating to in-vessel maintenance is currently under extensive development in accordance with the ITER R and D program. This paper summarizes the Japanese Home Team's contributions to remote maintenance design and technology development and outlines the plan on fabrication and testing of full-scale remote handling equipment/tools for in-vessel maintenance. (author)

  3. Localization of cask and plug remote handling system in ITER using multiple video cameras

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, João, E-mail: jftferreira@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Vale, Alberto [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building.

  4. Localization of cask and plug remote handling system in ITER using multiple video cameras

    International Nuclear Information System (INIS)

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building

  5. Remote handling concept for the neutral beam system

    International Nuclear Information System (INIS)

    The NB ITER Remote Maintenance System (NB IRMS) provides the means for the remote maintenance within the NB Cell by removal and replacement of the plant equipment. The NB IRMS will be installed and removed with the assistance of human workers during the preparation, and post-operation phase. During the maintenance operation after opening the Passive Magnetic Shield (PMS) and vessels, the maintenance activity and recovery from failure should be conducted remotely. This paper describes the concept design of the NB IRMS operating inside the NB cell for maintenance of the plant equipment such as NB components, and Upper Port Plugs (UPP). The main tasks of the IRMS, the description of the sub-systems and their specification, and deployment/operation principles are presented. The transportation concept of the NB IRMS to the hot cell facility for storage and maintenance is presented, which is to avoid unnecessary exposure on the equipment inside the NB cell during the machine operation.

  6. Preliminary definition of the remote handling system for the current IFMIF Test Facilities

    International Nuclear Information System (INIS)

    A coherent design of the remote handling system with the design of the components to be manipulated is vital for reliable, safe and fast maintenance, having a decisive impact on availability, occupational exposures and operational cost of the facility. Highly activated components in the IFMIF facility are found at the Test Cell, a shielded pit where the samples are accurately located. The remote handling system for the Test Cell reference design was outlined in some past IFMIF studies. Currently a new preliminary design of the Test Cell in the IFMIF facility is being developed, introducing important modifications with respect to the reference one. This recent design separates the previous Vertical Test Assemblies in three functional components: Test Modules, shielding plugs and conduits. Therefore, it is necessary to adapt the previous design of the remote handling system to the new maintenance procedures and requirements. This paper summarises such modifications of the remote handling system, in particular the assessment of the feasibility of a modified commercial multirope crane for the handling of the weighty shielding plugs for the new Test Cell and a quasi-commercial grapple for the handling of the new Test Modules.

  7. Remote waste handling at the Hot Fuel Examination Facility

    International Nuclear Information System (INIS)

    Radioactive solid wastes, some of which are combustible, are generated during disassembly and examination of irradiated fast-reactor fuel and material experiments at the Hot Fuel Examination Facility (HFEF). These wastes are remotely segregated and packaged in doubly contained, high-integrity, clean, retrievable waste packages for shipment to the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering Laboratory (INEL). This paper describes the equipment and techniques used to perform these operations

  8. A Feasible DEMO Blanket Concept Based on Water Cooled Solid Breeder

    International Nuclear Information System (INIS)

    Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

  9. Design and operation of a remotely operated plutonium waste size reduction and material handling process

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, III, J A; Charlesworth, D L

    1986-01-01

    Noncombustible /sup 238/Pu and /sup 239/Pu waste is generated as a result of normal operation and decommissioning activity at the Savannah River Plant, and is being retrievably stored there. As part of the long-term plant to process the stored waste and current waste for permanent disposal, a remote size reduction and material handling process is being cold-tested at Savannah River Laboratory. The process consists of a large, low-speed shredder and material handling system, a remote worktable, a bagless transfer system, and a robotically controlled manipulator. Initial testing of the shredder and material handling system and a cycle test of the bagless transfer system has been completed. Fabrication and acceptance testing of the Telerobat, a robotically controlled manipulator has been completed. Testing is scheduled to begin in 3/86. Design features maximizing the ability to remotely maintain the equipment were incorporated. Complete cold-testing of the equipment is scheduled to be completed in 1987.

  10. Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2009-10-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  11. Design and operation of a remotely operated plutonium waste size reduction and material handling process

    International Nuclear Information System (INIS)

    Non-combustible Pu-238 and Pu-239 waste is generated as a result of normal operation and decommissioning activity at the Savannah River Plant, and is being retrievably stored there. As part of the long-term plan to process the stored waste and current waste for permanent disposal, a remote size reduction and material handling process is being cold-tested at Savannah River Laboratory. The process consists of a large, low-speed shredder and material handling system, a remote worktable, a bagless transfer system, and a robotically controlled manipulator. Initial testing of the shredder and material handling system and a cycle test of the bagless transfer system has been completed. Fabrication and acceptance testing of the Telerobot, a robotically controlled manipulator, has been completed. Testing is scheduled to begin in 3/86. Design features maximizing the ability to remotely maintain the equipment were incorporated. Complete cold-testing of the equipment is scheduled to be completed in 987

  12. Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2011-03-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  13. Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2011-04-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  14. Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2010-06-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  15. Design and operation of a remotely operated plutonium waste size reduction and material handling process

    International Nuclear Information System (INIS)

    Noncombustible 238Pu and 239Pu waste is generated as a result of normal operation and decommissioning activity at the Savannah River Plant, and is being retrievably stored there. As part of the long-term plant to process the stored waste and current waste for permanent disposal, a remote size reduction and material handling process is being cold-tested at Savannah River Laboratory. The process consists of a large, low-speed shredder and material handling system, a remote worktable, a bagless transfer system, and a robotically controlled manipulator. Initial testing of the shredder and material handling system and a cycle test of the bagless transfer system has been completed. Fabrication and acceptance testing of the Telerobat, a robotically controlled manipulator has been completed. Testing is scheduled to begin in 3/86. Design features maximizing the ability to remotely maintain the equipment were incorporated. Complete cold-testing of the equipment is scheduled to be completed in 1987

  16. Reliability requirements management for ITER Remote Handling maintenance systems

    Energy Technology Data Exchange (ETDEWEB)

    Väyrynen, J., E-mail: jukka.vayrynen@tut.fi [Department of Intelligent Hydraulics and Automation, Tampere University of Technology, Tampere (Finland); Mattila, J. [Department of Intelligent Hydraulics and Automation, Tampere University of Technology, Tampere (Finland)

    2013-10-15

    Highlights: ► A model for reliability requirements management is presented. ► A proof of concept reliability allocation is made for a manipulator. ► System reliability and maintenance assessment is done based on the previous allocations. -- Abstract: This paper presents a model that can be used to ease reliability requirements management and designing reliability into the ITER remote maintenance equipment from the get-go, parallel to the mechanical design of the system. In addition, following the model presented automatically creates a reliability verification method during the physical design process with no or little additional cost and work involved.

  17. Remote handling equipment at the hanford waste treatment plant

    International Nuclear Information System (INIS)

    Cold war plutonium production led to extensive amounts of radioactive waste stored in tanks at the Department of Energy's Hanford Waste Treatment Plant. The storage tanks could potentially leak into the ground water and into the Columbia River. The solution for this risk of the leaking waste is vitrification. Vitrification is a process of mixing molten glass with radioactive waste to form a stable condition for storage. The Department of Energy has contracted Bechtel National, Inc. to build facilities at the Hanford site to process the waste. The waste will be separated into high and low level waste. Four major systems will process the waste, two pretreatment and two high level. Due to the high radiation levels, high integrity custom cranes have been designed to remotely maintain the hot cells. Several critical design parameters were implemented into the remote machinery design, including radiation limitations, remote operations, Important to Safety features, overall equipment effectiveness, minimum wall approaches, seismic constraints, and recovery requirements. Several key pieces of equipment were designed to meet these design requirements - high integrity crane bridges, trolleys, main hoists, mast hoists, slewing hoists, a monorail hoist, and telescoping mast deployed tele-robotic manipulator arms. There were unique and challenging design features and equipment needed to provide the remotely operated high integrity crane/manipulator systems for the Hanford Waste Treatment Plant. The cranes consist of a double girder bridge with various main hoist capacities ranging from one to thirty ton and are used for performing routine maintenance. A telescoping mast mounted tele-robotic manipulator arm with a one-ton hook is deployed from the trolley to perform miscellaneous operations in-cell. A dual two-ton slewing jib hoist is mounted to the bottom of the trolley and rotates 360 degrees around the mast allowing the closest hook wall approaches. Each of the two hoists on

  18. The integration of advanced technology: Robotics and remote handling

    International Nuclear Information System (INIS)

    Social, economic and environmental issues are placing increasing demands on the Nuclear Industry. There is a consequent need to develop new and existing technologies to respond, in a cost-effective manner, to these pressures. This paper deals with some of the specific, key, capabilities required in modern plants, with focus upon: (1) the ability to perform a wider range of operations in radioactive or toxic environments, automatically or remotely by intelligent robots; (2) the application and integration of advanced technology, including computer simulation and modelling, virtual reality, neural networks and expert systems. (author). 3 refs

  19. Reliability requirements management for ITER Remote Handling maintenance systems

    International Nuclear Information System (INIS)

    Highlights: ► A model for reliability requirements management is presented. ► A proof of concept reliability allocation is made for a manipulator. ► System reliability and maintenance assessment is done based on the previous allocations. -- Abstract: This paper presents a model that can be used to ease reliability requirements management and designing reliability into the ITER remote maintenance equipment from the get-go, parallel to the mechanical design of the system. In addition, following the model presented automatically creates a reliability verification method during the physical design process with no or little additional cost and work involved

  20. Factors affecting remote handling productivity during installation of the ITER-like wall at JET

    International Nuclear Information System (INIS)

    Highlights: ► The paper describes the challenges to achieve the installation of the ILW beryllium sliced wall. ► Examines the factual difference between estimated remote handling in-vessel durations and those achieved, with a view to quantifying the typical disparity between the two. ► The paper will elaborate and highlight the contributing factors. This offers an opportunity to provide provenance for availability estimates of devices such as ITER and DEMO. ► The paper will identify and describe the factors influencing the ratio between estimated versus the actual durations for remote handling operations. -- Abstract: Remote handling operations at JET have encountered many challenges to achieve the installation of the ILW beryllium sliced wall during the Enhanced Performance stage 2 (EP2) shutdown of JET. This was a demanding and challenging activity which was based on the experience gained from a period of over 15 years (20,000 h operations) of JET In-Vessel remote handling operations. This paper describes the difference between estimated remote handling in-vessel durations and those actually achieved with a view to quantifying the typical disparity between them. There are many factors that affect productivity of the remote handling operations and it is important to accommodate these either in the design of the component or within the production of the operational procedures with a view to minimise all impact on the final task duration. Some factors that affect the efficiency are outside the control of the design and operational procedures. These are unforeseen anomalies that were encountered during the removal, naked wall survey and installation of the components. Recoveries from these anomalies are extremely challenging and need to be addressed efficiently in order to minimise the impact on the shutdown duration and prevent optimised panned activities from becoming inefficient by fragmentation

  1. Radiation-tolerant cable management systems for remote handling applications in the nuclear industry

    International Nuclear Information System (INIS)

    Experience has shown that one of the most vulnerable areas within remote handling equipment is the umbilical cable and termination system. Repairs of a damaged system can be very long due to poorly designed termination techniques. Over the past five years W.L. Gore has gained considerable experience in the design and manufacture of cable systems, utilising unique radiation tolerant materials and manufacturing processes. The cable systems manufactured at the W.L. Gore, Dunfermline, Scotland facility have proven to give excellent performance in the most demanding of remote handling applications. (author)

  2. Defense remote-handled transuranic waste implementation plan: Transuranic Waste Program System Integration Office

    International Nuclear Information System (INIS)

    This document presents a detailed schedule for the implementation of the strategy for managing defense remote-handled (RH) transuranic (TRU) waste. The baseline management strategy was defined in the Defense Remote-Handled Transuranic Waste Cost/Schedule Optimization Study and is summarized in this document. Also included are revised RH TRU waste inventory projections, current site management plans, a list of key decision points and milestones, and a discussion of uncertainties associated with management of RH TRU waste. The plans are summarized in a detailed schedule diagram and in an RH TRU waste work off diagram. 9 refs., 5 figs., 4 tabs

  3. High-definition television evaluation for remote handling task performance

    International Nuclear Information System (INIS)

    This paper describes experiments designed to evaluate the impact of HDTV on the performance of typical remote tasks. The experiments described in this paper compared the performance of four operators using HDTV with their performance while using other television systems. The experiments included four television systems: (1) high-definition color television, (2) high-definition monochromatic television, (3) standard-resolution monochromatic television, and (4) standard-resolution stereoscopic monochromatic television. The stereo system accomplished stereoscopy by displaying two cross-polarized images, one reflected by a half-silvered mirror and one seen through the mirror. Observers wore a pair of glasses with cross-polarized lenses so that the left eye received only the view from the left camera and the right eye received only the view from the right camera

  4. Remote handling of TEXTOR diagnostics using CORBA as communication architecture

    International Nuclear Information System (INIS)

    At the Forschungszentrum Juelich, an upgrade of the existing distributed system for data acquisition (DAS) at the fusion experiment TEXTOR94 is under development. DAS is currently restricted to VAX/VMS and DECNET based communications, but it is planned to add UNIX based systems, and to open the local network for an improved wide area network access for remote operations. Therefore, the DAS system is to be equipped with a suitable client/server interface, which is able to cope with the various computer platforms and operating systems involved. For this purpose, the common object request broker architecture (CORBA) will be used. CORBA is an object oriented, standardized architecture for distributed systems, which provides a high degree of modularity in software design and allows for flexible implementations. It is to act as a connecting link between the existing system and new extensions. In order to provide the desired client/server functionality for the data acquisition tasks, the components of the system (diagnostic, database, etc.) are modelled by CORBA interfaces. Processes for diagnostic control and data readout in the existing OpenVMS systems are aimed at to be accessible by CORBA server implementations. The corresponding client implementations will be developed for the operating system platforms most frequently used at TEXTOR94. Communication between clients and server will be based on TCP/IP and are to be managed by CORBA. By this standardized way, remote control of diagnostic instrumentation becomes possible in a multiplatform computer and wide area network environment. At a later stage it is intended to integrate the system into a 'virtual control room' environment, which should enable the participation of cooperating institutions in the full experimental program of TEXTOR94. (orig.)

  5. Status of the extended performance tests for blanket remote maintenance in ITER L6 project

    International Nuclear Information System (INIS)

    Mechanically attached blanket module insertion tests were carried out considering the misalignment between module and back plate. Through the insertion tests, the module was successfully inserted up to the misalignment of ±10 mm under the clearance of ± 0.16 ∼ ±0.18 mm between key and groove. This was achieved by the passive compliance due to the flexibility of the manipulator through the assistance of the chamfer configuration of the key for smooth insertion. In addition, the 'correlation coefficient' based on the results obtained by the strain gages located at the end-effector was found to be useful in order to estimate the forces of the complicated end-effector during module insertion for the development of the sensor based control. (author)

  6. Status of extended performance tests for blanket remote maintenance in the ITER L6 project

    International Nuclear Information System (INIS)

    Mechanically attached blanket module insertion tests were carried out for various misalignments between the module and the back plate. In the insertion tests, the module was successfully inserted up to a misalignment of ±10 mm with a clearance of ±(0.16-0.18) mm between key and groove. This was achieved owing to the passive compliance due to the flexibility of the manipulator with the assistance of the chamfer configuration of the key for smooth insertion. In addition, the 'correlation coefficient' based on the results obtained by the strain gauges located at the end effector was found to be useful for estimating the forces of the complex end effector during module insertion for the development of sensor based control. (author)

  7. The functioning capability of a mobile remote handling system

    International Nuclear Information System (INIS)

    In areas exposed to ionizing radiation, for instance, in decommissioning and demolishing nuclear installations or in handling radioactive waste, so-called telerobotic systems are employed. These are integrated systems comprising a human operator and a robot with limited autonomy. The robots must be able to withstand radiation of up to 300 kGy. In order to make robot systems mobile, part of the electronics must be installed on board of these systems. Planning and design must bear in mind the high sensitivity to ionizing radiation of solid state components. The MF4 manipulator vehicle has been examined with respect to its sensitivity to ionizing radiation. The unit is a double track laying vehicle driven by electric motors and equipped with a manipulator, a stereoscopic camera, two cameras for driving, a dose rate probe, and a microphone. The service life of the vehicle has been extended considerably as a result of analyses of its individual components. A proposal is made in the report to shield extremely sensitive highly integrated subassemblies and make only enduring subassemblies of comparatively lower complexity more to radiation. In this way, tolerable radiation levels and the duration of missions can be increased. (orig.)

  8. An overview of the waste handling and packaging plant, a major processing facility for remote-handled transuranic waste

    International Nuclear Information System (INIS)

    The Waste Handling and Packaging Plant (WHPP) is a FY 1991 line item project proposed for construction at the Oak Ridge National Laboratory (ORNL). The purpose of the facility is to receive, package, certify and ship remote-handled (RH) and special case (SC) transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. The scope of the facility includes the mobilization of liquids and sludges from the Melton Valley Storage Tanks, transport of these liquids and sludges to the WHPP, solidification to a certifiable waste form, and final packaging and shipment to WIPP. Various solid hot cell wastes will be received at the WHPP from storage at ORNL and from other Department of Energy (DOE) sites. The solid wastes will be removed from the storage or shipping container, examined, processed as required, certified and packaged for shipment to WIPP. All packages coming from the processing cell will be in 55 gallon drums, and the facility will have the capability to load these directly into a shielded drum shipping cask, or to load these into the RH TRU canister for remote welding and shipment to WIPP using the RH TRU canister cask. 4 figs

  9. Evolving the JET virtual reality system for delivering the JET EP2 shutdown remote handling tasks

    International Nuclear Information System (INIS)

    The quality, functionality and performance of the virtual reality (VR) system used at JET for preparation and implementation of remote handling (RH) operations has been progressively enhanced since its first use in the original JET remote handling shutdown in 1998. As preparation began for the JET EP2 (Enhanced Performance 2) shutdown it was recognised that the VR system being used was unable to cope with the increased functionality and the large number of 3D models needed to fully represent the JET in-vessel components and tooling planned for EP2. A bespoke VR software application was developed in collaboration with the OEM, which allowed enhancements to be made to the VR system to meet the requirements of JET remote handling in preparation for EP2. Performance improvements required to meet the challenges of EP2 could not be obtained from the development of the new VR software alone. New methodologies were also required to prepare source, CATIA models for use in the VR using a collection of 3D software packages. In collaboration with the JET drawing office, techniques were developed within CATIA using polygon reduction tools to reduce model size, while retaining surface detail at required user limits. This paper will discuss how these developments have played an essential part in facilitating EP2 remote handling task development and examine their impact during the EP2 shutdown.

  10. Siting Study for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; Joan Connolly; Lance Peterson; Brennon Orr; Bob Starr

    2010-10-01

    The U.S. Department of Energy has identified a mission need for continued disposal capacity for remote-handled low-level waste (LLW) generated at the Idaho National Laboratory (INL). An alternatives analysis that was conducted to evaluate strategies to achieve this mission need identified two broad options for disposal of INL generated remote-handled LLW: (1) offsite disposal and (2) onsite disposal. The purpose of this study is to identify candidate sites or locations within INL boundaries for the alternative of an onsite remote handled LLW disposal facility and recommend the highest-ranked locations for consideration in the National Environmental Policy Act process. The study implements an evaluation based on consideration of five key elements: (1) regulations, (2) key assumptions, (3) conceptual design, (4) facility performance, and (5) previous INL siting study criteria, and uses a five-step process to identify, screen, evaluate, score, and rank 34 separate sites located across INL. The result of the evaluation is identification of two recommended alternative locations for siting an onsite remote-handled LLW disposal facility. The two alternative locations that best meet the evaluation criteria are (1) near the Advanced Test Reactor Complex and (2) west of the Idaho Comprehensive Environmental Response, Compensation, and Liability Act Disposal Facility.

  11. The remote handling compatibility analysis of the ITER generic upper port plug structure

    NARCIS (Netherlands)

    Ronden, D. M. S.; Dammann, A.; Elzendoorn, B.; Giacomin, T.; Heemskerk, C.; Loesser, D.; Maquet, P.; van Oosterhout, J.; Pak, S.; Pitcher, C. S.; M. Portalès,; Proust, M.; Udintsev, V.S.; Walsh, M. J.

    2014-01-01

    The ITER diagnostics generic upper port plug (GUPP) is developed as a standardized design for all diagnostic upper port plugs, in which a variety of payloads can be mounted. Here, the remote handling compatibility analysis (RHCA) of the GUPP design is presented that was performed for the GUPP final

  12. Robotics and remote handling concepts for disposal of high-level nuclear waste

    International Nuclear Information System (INIS)

    This paper summarizes preliminary remote handling and robotic concepts being developed as part of the US Department of Energy's (DOE) Yucca Mountain Project. The DOE is currently evaluating the Yucca Mountain Nevada site for suitability as a possible underground geologic repository for the disposal of high level nuclear waste. The current advanced conceptual design calls for the disposal of more than 12,000 high level nuclear waste packages within a 225 km underground network of tunnels and emplacement drifts. Many of the waste packages may weigh as much as 66 tonnes and measure 1.8 m in diameter and 5.6 m long. The waste packages will emit significant levels of radiation and heat. Therefore, remote handling is a cornerstone of the repository design and operating concepts. This paper discusses potential applications areas for robotics and remote handling technologies within the subsurface repository. It also summarizes the findings of a preliminary technology survey which reviewed available robotic and remote handling technologies developed within the nuclear, mining, rail and industrial robotics and automation industries, and at national laboratories, universities, and related research institutions and government agencies

  13. Remote technology related to the handling, storage and disposal of spent fuel. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    Reduced radiation exposure, greater reliability and cost savings are all potential benefits of the application of remote technologies to the handling of spent nuclear fuel. Remote equipment and technologies are used to some extent in all facilities handling fuel and high-level wastes whether they are for interim storage, processing/repacking, reprocessing or disposal. In view of the use and benefits of remote technologies, as well as recent technical and economic developments in the area, the IAEA organized the Technical Committee Meeting (TCM) on Remote Technology Related to the Handling, Storage and/or Disposal of Spent Fuel. Twenty-one papers were presented at the TCM, divided into five general areas: 1. Choice of technologies; 2. Use of remote technologies in fuel handling; 3. Use of remote technologies for fuel inspection and characterization; 4. Remote maintenance of facilities; and 5. Current and future developments. Refs, figs and tabs

  14. Remote handling in the Plutonium Immobilization Project: Puck packaging

    International Nuclear Information System (INIS)

    The Savannah River Site (SRS) will immobilize excess plutonium in the proposed Plutonium Immobilization Project (PIP). The PIP scope includes unloading transportation containers, preparing the feed streams, converting the metal feed to an oxide, adding the ceramic precursors, pressing the pucks, inspecting pucks, and sintering pucks. The PIP scope also includes loading the pucks into metal cans, sealing the cans, inspecting the cans, loading the cans into magazines, loading magazines into Defense Waste Processing Facility (DWPF) canisters, and transporting the canisters to the DWPF. The DWPF will fill the canister with a mixture of high-level waste and glass for permanent storage. Because of the radiation, remote equipment will perform PIP operations in a contained environment. The PIP puck packaging includes loading pucks into metal cans, sealing the cans, and inspecting the cans. A magnetically coupled elevator will lower a tray of pucks onto a magnetically coupled transport cart. This cart will carry the tray through an air lock into the can-loading glove box. Inside the glove box, a magnetically coupled tray lifter will raise the tray off the cart. A three-axis Cartesian robot will use a vacuum cup on a long pipe to lift the 67.3-mm (2.65-in.)-diam, 25.4-mm (1.0-in.)-tall pucks from the transfer tray and place 20 pucks in a 76.2-mm (3.0-in.)-diam stainless steel can. The Cartesian robot will place a custom hood on the open metal can, and this hood will remove the air from the can, insert helium, and place a hollow plug in the can. The SRS-developed bagless transfer system will weld the plug to the can wall and cut the can in the weld area. The can stub and the upper plug half above the cut line will remain in the sphincter seal to maintain the glove-box seal. The puck can and the lower plug half below the cut line is lowered into the bagless transfer enclosure. A floor-mounted robot in this enclosure will swipe the can exterior for contamination and place the

  15. Remote handling in highly radioactive environments and human intervention system in highly alpha contaminated environments

    International Nuclear Information System (INIS)

    Remote handling in highly contaminated environments is performed at this moment either with the help of mechanical manipulators connected to the biological protection, or with electromechanical remote manipulators carried on mobile positions covering the whole volume of the cell. The new electromechanical remote manipulator MA 23 M connected to a mobile positioner and to a servo TV system, allows to execute dexterity tasks with a minimum of fatigue for the operator; the total investment handling/vision being lower compared with the use of mechanical manipulators and vision windows, more specially when using the manipulators for maintenance. The intervention system SCALHENE, specific application of the double door transfer system DPTE type, allows a man to be introduced directly into a hostile environment without having to pass through an intermediate chamber

  16. Calculations to support JET neutron yield calibration: Modelling of the JET remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka, E-mail: luka.snoj@ijs.si [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-MHEST Association, Reactor Physics Division, Jožef Stefan Institute, Jamova Cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Čufar, Aljaž [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-MHEST Association, Reactor Physics Division, Jožef Stefan Institute, Jamova Cesta 39, SI-1000 Ljubljana (Slovenia); Syme, Brian; Popovichev, Sergey [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB, OXON (United Kingdom); Conroy, Sean [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-VR Association, Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Meredith, Lewis [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB, OXON (United Kingdom)

    2013-08-15

    Highlights: ► We model JET remote handling system in MCNP. ► We examine the effect of JET remote handling system on neutron monitor response. ► The integral effect of JET RH system on neutron monitors is less than 5%. -- Abstract: After the coated CFC wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) transition in 2010–2011, confirmation of the neutron yield calibration will be ensured by direct measurements using a calibrated {sup 252}Cf neutron source deployed by the in-vessel remote handling boom and Mascot manipulator inside the JET vacuum vessel. Neutronic calculations are required to calculate the effects of the JET remote handling (RH) system on the neutron monitors. We developed a simplified geometrical computational model of the JET remote handling system in MCNP. In parallel we developed a script that translates the RH movement data to transformations of individual geometrical parts of the RH model in MCNP. After that a benchmarking of the model was performed to verify and validate the accordance of the target positions of source and RH system with the ones from our model. In the last phase we placed the JET RH system in the simplified MCNP model of the JET tokamak and studied its effect on neutron monitor response for some example source positions and boom configurations. As the correction factors due to presence of the JET RH system can potentially be significant in cases when the boom is blocking a port close to the detector under investigation, we have chosen boom configurations so that this is avoided in the vast majority of the source locations. Examples are given.

  17. Unique Remote-Handled Waste Management Issues at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Since the Manhattan Project, Oak Ridge National Laboratory (ORNL) has been engaged in developing processes for implementation in the Department of Energy (DOE) production facilities and in producing radioisotopes for medical and industrial applications. These activities have resulted in a large variety of unique remote-handled legacy waste and contaminated hot cell facilities. The DOE has established the Integrated Facility Disposition Project (IFDP) to dispose of the ORNL legacy waste and to deactivate, decontaminate, and decommission facilities at ORNL no longer needed for the mission. The IFDP will be required to characterize, treat, package, and dispose of various remote-handled solid waste streams for which no treatment capability currently exists at ORNL. This paper describes the capabilities that will be required to manage these waste streams and the options evaluated for implementation. The IFDP CD-1 reflects the construction of a remote-handled solids-processing facility designed to meet current safety requirements to ensure the safety of workers and the public and to include all the necessary capabilities to handle the suite of IFDP wastes that have been identified. The new solids-processing facility will be designed so that it can be transitioned to ORNL at the end of the IFDP and used for long-term management of waste from ongoing research missions. The single new facility constructed to handle all remote-handled solid waste streams from IFDP was the only one that provides the capabilities to process the full suite of IFDP materials without major modifications to existing nuclear facilities. To support the CD-2 development, an additional evaluation will be performed to determine if a less expensive alternative can be identified by using a combination of existing facilities to treat subsets of waste streams. (authors)

  18. High Level Waste Remote Handling Equipment in the Melter Cave Support Handling System at the Hanford Waste Treatment Plant

    International Nuclear Information System (INIS)

    Cold war plutonium production led to extensive amounts of radioactive waste stored in tanks at the Department of Energy's (DOE) Hanford site. Bechtel National, Inc. is building the largest nuclear Waste Treatment Plant in the world located at the Department of Energy's Hanford site to immobilize the millions of gallons of radioactive waste. The site comprises five main facilities; Pretreatment, High Level Waste vitrification, Low Active Waste vitrification, an Analytical Lab and the Balance of Facilities. The pretreatment facilities will separate the high and low level waste. The high level waste will then proceed to the HLW facility for vitrification. Vitrification is a process of utilizing a melter to mix molten glass with radioactive waste to form a stable product for storage. The melter cave is designated as the High Level Waste Melter Cave Support Handling System (HSH). There are several key processes that occur in the HSH cell that are necessary for vitrification and include: feed preparation, mixing, pouring, cooling and all maintenance and repair of the process equipment. Due to the cell's high level radiation, remote handling equipment provided by PaR Systems, Inc. is required to install and remove all equipment in the HSH cell. The remote handling crane is composed of a bridge and trolley. The trolley supports a telescoping tube set that rigidly deploys a TR 4350 manipulator arm with seven degrees of freedom. A rotating, extending, and retracting slewing hoist is mounted to the bottom of the trolley and is centered about the telescoping tube set. Both the manipulator and slewer are unique to this cell. The slewer can reach into corners and the manipulator's cross pivoting wrist provides better operational dexterity and camera viewing angles at the end of the arm. Since the crane functions will be operated remotely, the entire cell and crane have been modeled with 3-D software. Model simulations have been used to confirm operational and maintenance

  19. Analysis of operational possibilities and conditions of remote handling systems in nuclear facilities

    International Nuclear Information System (INIS)

    Accepting the development of the occupational radiation exposure in nuclear facilities, it will be showing possibilities of cost effective reduction of the dose rate through the application of robots and manipulators for the maintenance of nuclear power plants, fuel reprocessing plants, decommissioning and dismantling of the mentioned plants. Based on the experiences about industrial robot applications by manufacturing and manipulator applications by the handling of radioactive materials as well as analysis of the handling procedures and estimation of the dose intensity, it will be defining task-orientated requirements for the conceptual design of the remote handling systems. Furthermore the manifold applications of stationary and mobil arranged handling systems in temporary or permanent operation are described. (orig.)

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

    International Nuclear Information System (INIS)

    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

  1. Flexible path optimization for the Cask and Plug Remote Handling System in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Vale, Alberto, E-mail: avale@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Fonte, Daniel; Valente, Filipe; Ferreira, João [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Gonzalez, Carmen [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: ► Complementary approach for path optimization named free roaming that takes full advantage of the rhombic like kinematics of the Cask and Plug Remote Handling System (CPRHS). ► Possibility to find trajectories not possible in the past using the line guidance developed in a previous work, in particular when moving the Cask Transfer System (CTS) beneath the pallet or in rescue missions. ► Methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. -- Abstract: The Cask and Plug Remote Handling System (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell Building and the Tokamak Building in ITER along pre-defined optimized trajectories. A first approach for CPRHS path optimization was previously proposed using line guidance as the navigation methodology to be adopted. This approach might not lead to feasible paths in new situations not considered during the previous work, as rescue operations. This paper addresses this problem by presenting a complementary approach for path optimization inspired in rigid body dynamics that takes full advantage of the rhombic like kinematics of the CPRHS. It also presents a methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. The results gathered from 500 optimized trajectories are summarized. Conclusions and open issues are presented and discussed.

  2. Flexible path optimization for the Cask and Plug Remote Handling System in ITER

    International Nuclear Information System (INIS)

    Highlights: ► Complementary approach for path optimization named free roaming that takes full advantage of the rhombic like kinematics of the Cask and Plug Remote Handling System (CPRHS). ► Possibility to find trajectories not possible in the past using the line guidance developed in a previous work, in particular when moving the Cask Transfer System (CTS) beneath the pallet or in rescue missions. ► Methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. -- Abstract: The Cask and Plug Remote Handling System (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell Building and the Tokamak Building in ITER along pre-defined optimized trajectories. A first approach for CPRHS path optimization was previously proposed using line guidance as the navigation methodology to be adopted. This approach might not lead to feasible paths in new situations not considered during the previous work, as rescue operations. This paper addresses this problem by presenting a complementary approach for path optimization inspired in rigid body dynamics that takes full advantage of the rhombic like kinematics of the CPRHS. It also presents a methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. The results gathered from 500 optimized trajectories are summarized. Conclusions and open issues are presented and discussed

  3. Demonstration of remotely operated TRU waste size reduction and material handling equipment

    Energy Technology Data Exchange (ETDEWEB)

    Looper, M G; Charlesworth, D L

    1988-01-01

    The Savannah River Laboratory (SRL) is developing remote size reduction and material handling equipment to prepare /sup 238/Pu contaminated waste for permanent disposal at the Waste Isolation Pilot Plant (WIPP) in New Mexico. The waste is generated at the Savannah River Plant (SRP) from normal operation and decommissioning activity and is retrievably stored onsite. A Transuranic Waste Facility for preparing, size-reducing, and packaging this waste for disposal is scheduled for completion in 1995. A cold test facility for demonstrating the size reduction and material handling equipment was built, and testing began in January 1987. 9 figs., 1 tab.

  4. Demonstration of remotely operated TRU waste size reduction and material handling equipment

    International Nuclear Information System (INIS)

    The Savannah River Laboratory (SRL) is developing remote size reduction and material handling equipment to prepare 238Pu contaminated waste for permanent disposal at the Waste Isolation Pilot Plant (WIPP) in New Mexico. The waste is generated at the Savannah River Plant (SRP) from normal operation and decommissioning activity and is retrievably stored onsite. A Transuranic Waste Facility for preparing, size-reducing, and packaging this waste for disposal is scheduled for completion in 1995. A cold test facility for demonstrating the size reduction and material handling equipment was built, and testing began in January 1987. 9 figs., 1 tab

  5. Distance sensing for the remote handling equipment with radiation resistant ultrasonic transducers up to 100 MGy

    International Nuclear Information System (INIS)

    Ultrasonic sensors provide a good way to measure distances and avoid collision on the remote handling units used for the fusion reactors maintenance. This paper describes the results of on-going irradiation tests performed on chosen ultrasonic sensors up to a total dose of 100 MGy. The present results at 92 MGy clearly show that although the sensitivity of the sensors degrade significantly and that their operating resonance frequency changes with increasing total dose, their operability is maintained. A dedicated electronic circuit has been developed to operate them remotely with longer cables than the standard ones. (authors)

  6. Interim design status and operational report for remote handling fixtures: primary and secondary burners

    International Nuclear Information System (INIS)

    The HTGR reprocessing flowsheet consists of two basic process elements: (1) spent fuel crushing and burning and (2) solvent extraction. Fundamental to these elements is the design and development of specialized process equipment and support facilities. A major consideration of this design and development program is equipment maintenance: specifically, the design and demonstration of selected remote maintenance capabilities and the integration of these into process equipment design. This report documents the current status of the development of remote handling and maintenance fixtures for the primary and secondary burners

  7. Overview of Remote Handling Equipment Used for the NPP A1 Decommissioning - 12141

    International Nuclear Information System (INIS)

    The first Czechoslovak NPP A1 was in operation from 1972 to 1977 and it was finally shutdown due to an accident (level 4 according to the INES). The presence of radioactive, toxic or hazardous materials limits personnel access to facilities and therefore it is necessary to use remote handling technologies for some most difficult characterization, retrieval, decontamination and dismantling tasks. The history of remote handling technologies utilization started in nineties when the spent nuclear fuel, including those fuel assemblies damaged during the accident, was prepared for the transport to Russia. Subsequent significant development of remote handling equipment continued during implementation of the NPP A1 decommissioning project - Stage I and ongoing Stage II. Company VUJE, Inc. is the general contractor for both mentioned stages of the decommissioning project. Various remote handling manipulators and robotics arms were developed and used. It includes remotely controlled vehicle manipulator MT-15 used for characterisation tasks in hostile and radioactive environment, special robust manipulator DENAR-41 used for the decontamination of underground storage tanks and multi-purposes robotics arms MT-80 and MT-80A developed for variety of decontamination and dismantling tasks. The heavy water evaporator facility dismantling is the current task performed remotely by robotics arm MT-80. The heavy water evaporator is located inside the main production building in the room No. 220 where loose surface contamination varies from 10 Bq/cm2 to 1x103 Bq/cm2, dose rate is up to 1.5 mGy/h and the feeding pipeline contained liquid RAW with high tritium content. Presented manipulators have been designed for broad range of decommissioning tasks. They are used for recognition, sampling, waste retrieval from large underground tanks, decontamination and dismantling of technological equipments. Each of the mentioned fields claims specific requirements on design of manipulator, their

  8. On-site transfer system for remote handling of low-level radioactive waste

    International Nuclear Information System (INIS)

    Increased uncertainties regarding the future availability of low-level radioactive waste (LLW) disposal sites have caused many commercial nuclear power utilities to investigate and implement alternatives to radwaste storage and disposal. Nuclear Packaging, Inc., under contract to Southern California Edison has developed an on-site radioactive waste transfer system (OTS), which allows shielded handling of LLW at the San Onofre Nuclear Generating Station. The system is designed to remotely transfer multiconfigured radwaste containers into shielded storage modules, on-site radioactive waste storage facilities, or shipping casks. The OTS consists of three primary components: (a) a shielded transfer cask, (b) a transport trailer, and (c) a mobile straddle crane for remote handling and positioning of the transfer cask during container transfer operations

  9. Conceptual Design Report for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; David Duncan; Joan Connolly; Margaret Hinman; Charles Marcinkiewicz; Gary Mecham

    2011-03-01

    This conceptual design report addresses development of replacement remote-handled low-level waste disposal capability for the Idaho National Laboratory. Current disposal capability at the Radioactive Waste Management Complex is planned until the facility is full or until it must be closed in preparation for final remediation (approximately at the end of Fiscal Year 2017). This conceptual design report includes key project assumptions; design options considered in development of the proposed onsite disposal facility (the highest ranked alternative for providing continued uninterrupted remote-handled low level waste disposal capability); process and facility descriptions; safety and environmental requirements that would apply to the proposed facility; and the proposed cost and schedule for funding, design, construction, and operation of the proposed onsite disposal facility.

  10. Conceptual Design Report for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2011-05-01

    This conceptual design report addresses development of replacement remote-handled low-level waste disposal capability for the Idaho National Laboratory. Current disposal capability at the Radioactive Waste Management Complex is planned until the facility is full or until it must be closed in preparation for final remediation (approximately at the end of Fiscal Year 2017). This conceptual design report includes key project assumptions; design options considered in development of the proposed onsite disposal facility (the highest ranked alternative for providing continued uninterrupted remote-handled low level waste disposal capability); process and facility descriptions; safety and environmental requirements that would apply to the proposed facility; and the proposed cost and schedule for funding, design, construction, and operation of the proposed onsite disposal facility.

  11. Conceptual design for remote handling methods using the HIP process in the Calcine Immobilization Program

    International Nuclear Information System (INIS)

    This report recommends the remote conceptual design philosophy for calcine immobilization using the hot isostatic press (HIP) process. Areas of remote handling operations discussed in this report include: (1) introducing the process can into the front end of the HIP process, (2) filling and compacting the calcine/frit mixture into the process can, (3) evacuating and sealing the process can, (4) non-destructive testing of the seal on the process can, (5) decontamination of the process can, (6) HIP furnace loading and unloading the process can for the HIPing operation, (7) loading an overpack canister with processed HIP cans, (8) sealing the canister, with associated non-destructive examination (NDE) and decontamination, and (9) handling canisters for interim storage at the Idaho Chemical Processing Plant (ICPP) located on the Idaho National Engineering Laboratory (INEL) site

  12. Use of mobile remote-handling equipment in connection with nuclear accidents

    International Nuclear Information System (INIS)

    In the event of a nuclear accident it will be necessary to start operations as soon as possible in zones inaccessible to man in order to determine the nature and extent of damage (measurement of various parameters, visual observation and photography) and to take action with a view to limiting the consequences of the accident. These various operations will have to be performed by remote-handling equipment containing measurement devices, television cameras and manipulators. The development of devices for immediate action is not independent of that of devices for subsequent operations either for repair or for dismantling. The paper describes the devices used at present in France and reports the experience gained over several years by the CEA Mobile Remote-Handling Team during action at nuclear facilities. It suggests a strategy for the utilization of such equipment and outlines a programme for developing essential facilities, taking into account the possibilities of international assistance. (author)

  13. Conceptual Design Report for Remote-Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; David Duncan; Joan Connolly; Margaret Hinman; Charles Marcinkiewicz; Gary Mecham

    2010-10-01

    This conceptual design report addresses development of replacement remote-handled low-level waste disposal capability for the Idaho National Laboratory. Current disposal capability at the Radioactive Waste Management Complex is planned until the facility is full or until it must be closed in preparation for final remediation (approximately at the end of Fiscal Year 2017). This conceptual design report includes key project assumptions; design options considered in development of the proposed onsite disposal facility (the highest ranked alternative for providing continued uninterrupted remote-handled low level waste disposal capability); process and facility descriptions; safety and environmental requirements that would apply to the proposed facility; and the proposed cost and schedule for funding, design, construction, and operation of the proposed onsite disposal facility.

  14. Evaluation of a New Remote Handling Design for High Throughput Annular Centrifugal Contactors

    Energy Technology Data Exchange (ETDEWEB)

    David H. Meikrantz; Troy G. Garn; Jack D. Law; Lawrence L. Macaluso

    2009-09-01

    Advanced designs of nuclear fuel recycling plants are expected to include more ambitious goals for aqueous based separations including; higher separations efficiency, high-level waste minimization, and a greater focus on continuous processes to minimize cost and footprint. Therefore, Annular Centrifugal Contactors (ACCs) are destined to play a more important role for such future processing schemes. Previous efforts defined and characterized the performance of commercial 5 cm and 12.5 cm single-stage ACCs in a “cold” environment. The next logical step, the design and evaluation of remote capable pilot scale ACCs in a “hot” or radioactive environment was reported earlier. This report includes the development of remote designs for ACCs that can process the large throughput rates needed in future nuclear fuel recycling plants. Novel designs were developed for the remote interconnection of contactor units, clean-in-place and drain connections, and a new solids removal collection chamber. A three stage, 12.5 cm diameter rotor module has been constructed and evaluated for operational function and remote handling in highly radioactive environments. This design is scalable to commercial CINC ACC models from V-05 to V-20 with total throughput rates ranging from 20 to 650 liters per minute. The V-05R three stage prototype was manufactured by the commercial vendor for ACCs in the U.S., CINC mfg. It employs three standard V-05 clean-in-place (CIP) units modified for remote service and replacement via new methods of connection for solution inlets, outlets, drain and CIP. Hydraulic testing and functional checks were successfully conducted and then the prototype was evaluated for remote handling and maintenance suitability. Removal and replacement of the center position V-05R ACC unit in the three stage prototype was demonstrated using an overhead rail mounted PaR manipulator. This evaluation confirmed the efficacy of this innovative design for interconnecting and cleaning

  15. Evaluation of a New Remote Handling Design for High Throughput Annular Centrifugal Contactors

    International Nuclear Information System (INIS)

    Advanced designs of nuclear fuel recycling plants are expected to include more ambitious goals for aqueous based separations including; higher separations efficiency, high-level waste minimization, and a greater focus on continuous processes to minimize cost and footprint. Therefore, Annular Centrifugal Contactors (ACCs) are destined to play a more important role for such future processing schemes. Previous efforts defined and characterized the performance of commercial 5 cm and 12.5 cm single-stage ACCs in a 'cold' environment. The next logical step, the design and evaluation of remote capable pilot scale ACCs in a 'hot' or radioactive environment was reported earlier. This report includes the development of remote designs for ACCs that can process the large throughput rates needed in future nuclear fuel recycling plants. Novel designs were developed for the remote interconnection of contactor units, clean-in-place and drain connections, and a new solids removal collection chamber. A three stage, 12.5 cm diameter rotor module has been constructed and evaluated for operational function and remote handling in highly radioactive environments. This design is scalable to commercial CINC ACC models from V-05 to V-20 with total throughput rates ranging from 20 to 650 liters per minute. The V-05R three stage prototype was manufactured by the commercial vendor for ACCs in the U.S., CINC mfg. It employs three standard V-05 clean-in-place (CIP) units modified for remote service and replacement via new methods of connection for solution inlets, outlets, drain and CIP. Hydraulic testing and functional checks were successfully conducted and then the prototype was evaluated for remote handling and maintenance suitability. Removal and replacement of the center position V-05R ACC unit in the three stage prototype was demonstrated using an overhead rail mounted PaR manipulator. This evaluation confirmed the efficacy of this innovative design for interconnecting and cleaning

  16. The development and evaluation of a stereoscopic television system for remote handling

    International Nuclear Information System (INIS)

    This paper describes the development and evaluation of a stereoscopic television system at Harwell Laboratory. The theory of stereo image geometry is outlined, and criteria for the matching of stereoscopic pictures are given. A stereoscopic television system designed for remote handling tasks has been produced, it provides two selectable angles of view and variable convergence, the display is viewed via polarizing spectacles. Evaluations have indicated improved performance with no problems of operator fatigue over a wide range of applications. (author)

  17. Remote Handled Transuranic Sludge Retrieval Transfer And Storage System At Hanford

    International Nuclear Information System (INIS)

    This paper describes the systems developed for processing and interim storage of the sludge managed as remote-handled transuranic (RH-TRU). An experienced, integrated CH2M HILL/AFS team was formed to design and build systems to retrieve, interim store, and treat for disposal the K West Basin sludge, namely the Sludge Treatment Project (STP). A system has been designed and is being constructed for retrieval and interim storage, namely the Engineered Container Retrieval, Transfer and Storage System (ECRTS)

  18. Acquisition Strategy for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposition Project

    International Nuclear Information System (INIS)

    This document describes the design-build acquisition strategy that will be applied to the Remote Handled LLW Disposal Project. The design-build delivery method will be tailored, as appropriate, to integrate the requirements of Department of Energy (DOE) Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets,' with the DOE budget formulation process and the safety requirements of DOE-STD-1189, 'Integration of Safety into the Design Process.'

  19. A passive-active neutron device for assaying remote-handled transuranic waste

    International Nuclear Information System (INIS)

    A combined passive-active neutron assay device was constructed for assaying remote-handled transuranic waste. A study of matrix and source position effects in active assays showed that a knowledge of the source position alone is not sufficient to correct for position-related errors in highly moderating or absorbing matrices. An alternate function for the active assay of solid fuel pellets was derived, although the efficacy of this approach remains to be established

  20. Benchmarking the Remote-Handled Waste Facility at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    O. P. Mendiratta; D. K. Ploetz

    2000-02-29

    ABSTRACT Facility decontamination activities at the West Valley Demonstration Project (WVDP), the site of a former commercial nuclear spent fuel reprocessing facility near Buffalo, New York, have resulted in the removal of radioactive waste. Due to high dose and/or high contamination levels of this waste, it needs to be handled remotely for processing and repackaging into transport/disposal-ready containers. An initial conceptual design for a Remote-Handled Waste Facility (RHWF), completed in June 1998, was estimated to cost $55 million and take 11 years to process the waste. Benchmarking the RHWF with other facilities around the world, completed in November 1998, identified unique facility design features and innovative waste pro-cessing methods. Incorporation of the benchmarking effort has led to a smaller yet fully functional, $31 million facility. To distinguish it from the June 1998 version, the revised design is called the Rescoped Remote-Handled Waste Facility (RRHWF) in this topical report. The conceptual design for the RRHWF was completed in June 1999. A design-build contract was approved by the Department of Energy in September 1999.

  1. Benchmarking the Remote-Handled Waste Facility at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    ABSTRACT Facility decontamination activities at the West Valley Demonstration Project (WVDP), the site of a former commercial nuclear spent fuel reprocessing facility near Buffalo, New York, have resulted in the removal of radioactive waste. Due to high dose and/or high contamination levels of this waste, it needs to be handled remotely for processing and repackaging into transport/disposal-ready containers. An initial conceptual design for a Remote-Handled Waste Facility (RHWF), completed in June 1998, was estimated to cost $55 million and take 11 years to process the waste. Benchmarking the RHWF with other facilities around the world, completed in November 1998, identified unique facility design features and innovative waste processing methods. Incorporation of the benchmarking effort has led to a smaller yet fully functional, $31 million facility. To distinguish it from the June 1998 version, the revised design is called the Rescoped Remote-Handled Waste Facility (RRHWF) in this topical report. The conceptual design for the RRHWF was completed in June 1999. A design-build contract was approved by the Department of Energy in September 1999

  2. Measurement and control system for ITER remote maintenance equipment

    International Nuclear Information System (INIS)

    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)

  3. Augmented virtualised reality-Applications and benefits in remote handling for fusion

    International Nuclear Information System (INIS)

    Over the last 10 years VR has been used at JET in an increasingly important role. It now finds use in various aspects of task preparation including planning, mock-up, training and task overview. It also plays an important role in actual operations where it is used to gain a more complete view of the work area. The JET VR implementation does not have on-line monitoring of the remote environment and the robot modelling has accuracy limitations, so this system cannot be used as the primary means of viewing. Work is currently underway with the aim of allowing such as system to run at ITER with full remote environment monitoring with high enough precision and accuracy so as to allow its use as the primary viewing method. This paper looks at how this augmented virtualised reality solution would be applied and considers some of the additional benefits AVR could have in remote handling for fusion.

  4. Mission Need Statement for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego

    2009-06-01

    The Idaho National Laboratory proposes to establish replacement remote-handled low-level waste disposal capability to meet Nuclear Energy and Naval Reactors mission-critical, remote-handled low-level waste disposal needs beyond planned cessation of existing disposal capability at the end of Fiscal Year 2015. Remote-handled low-level waste is generated from nuclear programs conducted at the Idaho National Laboratory, including spent nuclear fuel handling and operations at the Naval Reactors Facility and operations at the Advanced Test Reactor. Remote-handled low-level waste also will be generated by new programs and from segregation and treatment (as necessary) of remote-handled scrap and waste currently stored in the Radioactive Scrap and Waste Facility at the Materials and Fuels Complex. Replacement disposal capability must be in place by Fiscal Year 2016 to support uninterrupted Idaho operations. This mission need statement provides the basis for the laboratory’s recommendation to the Department of Energy to proceed with establishing the replacement remote-handled low-level waste disposal capability, project assumptions and constraints, and preliminary cost and schedule information for developing the proposed capability. Without continued remote-handled low-level waste disposal capability, Department of Energy missions at the Idaho National Laboratory would be jeopardized, including operations at the Naval Reactors Facility that are critical to effective execution of the Naval Nuclear Propulsion Program and national security. Remote-handled low-level waste disposal capability is also critical to the Department of Energy’s ability to meet obligations with the State of Idaho.

  5. New System For Tokamak T-10 Experimental Data Acquisition, Data Handling And Remote Access

    International Nuclear Information System (INIS)

    For carrying out the experiments on nuclear fusion devices in the Institute of Nuclear Fusion, Moscow, a system for experimental data acquisition, data handling and remote access (further 'DAS-T10') was developed and has been used in the Institute since the year 2000. The DAS-T10 maintains the whole cycle of experimental data handling: from configuration of data measuring equipment and acquisition of raw data from the fusion device (the Device), to presentation of math-processed data and support of the experiment data archive. The DAS-T10 provides facilities for the researchers to access the data both at early stages of an experiment and well afterwards, locally from within the experiment network and remotely over the Internet.The DAS-T10 is undergoing a modernization since the year 2007. The new version of the DAS-T10 will accommodate to modern data measuring equipment and will implement improved architectural solutions. The innovations will allow the DAS-T10 to produce and handle larger amounts of experimental data, thus providing the opportunities to intensify and extend the fusion researches. The new features of the DAS-T10 along with the existing design principles are reviewed in this paper

  6. Demonstration of a remotely operated TRU waste size-reduction and material handling process

    International Nuclear Information System (INIS)

    Noncombustible Pu-238 and Pu-239 waste is generated as a result of normal operation and decommissioning activity at the Savannah River Plant and is being retrievably stored at the site. As part of the long-term plan to process the stored waste and current waste for permanent disposal, a remote size-reduction and material handling process is being tested at Savannah River Laboratory to provide design support for the plant TRU Waste Facility scheduled to be completed in 1993. The process consists of a large, low-speed shredder and material handling system, a remote worktable, a bagless transfer system, and a robotically controlled manipulator, or Telerobot. Initial testing of the shredder and material handling system and a cycle test of the bagless transfer system were completed. Initial Telerobot run-in and system evaluation was completed. User software was evaluated and modified to support complete menu-driven operation. Telerobot prototype size-reduction tooling was designed and successfully tested. Complete nonradioactive testing of the equipment is scheduled to be completed in 1987

  7. Failure Mode and Effect Analysis for remote handling transfer systems of ITER

    International Nuclear Information System (INIS)

    A Failure Mode and Effect Analysis (FMEA) at component level was done to study safety-relevant implications arising from possible failures in performing remote handling (RH) operations at ITER facility . Autonomous air cushion transporter, pallet, sealed casks and tractor movers needed for port plug mounting/dismantling operation were analysed. For each sub-system, the breakdown of significant components was outlined and, for each component, possible failure modes have been investigated pointing out possible causes, possible actions to prevent the causes, consequences and actions to prevent or mitigate consequences. Off-normal events which may result in hazardous consequences to the public and the environment have been defined as Postulated Initiating Events (PIEs). Two safety-relevant PIEs have been defined by assessing elementary failures related to the analysed system. Each PIE has been discussed in order to qualitatively identify accident sequences arising from each of them. As an output of this FMEA study, possible incidental scenarios, where the intervention of rescue RH equipments is required to overcome critical situations determined by fault of RH components, were defined as well. Being rescue scenarios of main concern for ITER remote handling activities, such families could be helpful in defining the design requirements of port handling systems in general and on RH transfer system in particular. Furthermore, they could be useful in defining casks and vehicles to be used for rescue activities

  8. Integrated digital control and man-machine interface for complex remote handling systems

    International Nuclear Information System (INIS)

    The Advanced Integrated Maintenance System (AIMS) is part of a continuing effort within the Consolidated Fuel Reprocessing Program at Oak Ridge National Laboratory to develop and extend the capabilities of remote manipulation and maintenance technology. The AIMS is a totally integrated approach to remote handling in hazardous environments. State-of-the-art computer systems connected through a high-speed communication network provide a real-time distributed control system that supports the flexibility and expandability needed for large integrated maintenance applications. A Man-Machine Interface provides high-level human interaction through a powerful color graphics menu-controlled operator console. An auxiliary control system handles the real-time processing needs for a variety of support hardware. A pair of dedicated fiber-optic-linked master/slave computer system control the Advanced Servomanipulator master/slave arms using powerful distributed digital processing methods. The FORTH language was used as a real-time operating and development environment for the entire system, and all of these components are integrated into a control room concept that represents the latest advancements in the development of remote maintenance facilities for hazardous environments

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

    International Nuclear Information System (INIS)

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

  10. State and outlooks of remote handling and automation techniques use for industrial radioactive operations

    International Nuclear Information System (INIS)

    Handling in reactors mainly concerns charging and discharging operations and inspection. Specific means are being developed for each operation, with an increasing degree of automation. This serves to reduce exposure of personnel. However, the development of these means conflicts in certain cases with the original plant design, which did not provide for remote maintenance. With regard to fuel reprocessing, handling at the processing level is becoming increasingly automated. The difficulties lie principally in maintenance and waste conditioning operations. These involve less specialized means than is the case with reactors and can only be automated to a limited extent, save in exceptional cases. The greatest progress will be achieved by laying down stringent maintenance principles and taking them into consideration at the design stage

  11. Remote-handling devices for radioactive materials - Part 2: Mechanical master-slave manipulators

    International Nuclear Information System (INIS)

    ISO 17874 consists of the following parts, under the general title Remote-handling devices for radioactive materials: Part 1: General requirements; Part 2: Mechanical master-slave manipulators; Part 3: Electrical master-slave manipulators; Part 4: Power manipulators; Part 5: Remote-handling tongs. This part of ISO 17874 deals with mechanical master-slave manipulators used for nuclear applications. These devices replace the hands and arms of the operators in areas inaccessible to personnel (mostly behind shielding walls). Mechanical master-slave manipulators were originally developed for hot cells, which were designed for research and development for nuclear power reactor-fuel elements. They are now also in use in other nuclear installations, such as fabrication or reprocessing plants for fuel elements, waste-treatment stations and decommissioning of nuclear facilities. This part of ISO 17874 should be of assistance to designers of nuclear plants, as well as to manufacturers, users and license authorities. This part (2) of ISO 17874 specifies the criteria for the selection, installation and use of a mechanical master-slave manipulator, for remote handling of radiaoactive materials in a nuclear facility. This part of ISO 17874 deals only with the technical aspects related to the manipulator and its interface with the nuclear facility in which it is intended to be installed. In particular, the process apparatus and the manipulator features need to be studied in parallel in order to optimize all the functionalities of the manipulator. However, this part of ISO 17874 does not cover the fundamental design criteria of the nuclear facility (e.g. the process involved, maintenance of the process equipment, intervention for other purposes)

  12. ITER ECH and CD Upper Launcher: Design options and Remote Handling issues of the waveguide assembly

    International Nuclear Information System (INIS)

    Highlights: • Paper deals with ITER Electron Cyclotron Heating and Current Drive Upper Launcher. • Design options and Remote Handling issues related to the waveguide assembly have been investigated. • Description of preliminary assessments on the RH compatibility of the sub-assembly is shown. • Assessment on possible replacement procedure and the required tools are presented. - Abstract: The ITER Electron Cyclotron Heating and Current Drive Upper Launcher, developed by the ECHUL-CA Consortium of Euratom Associations (CRPP, CNR, ITER-NL, IPF, IPP, KIT and Politecnico di Milano), is presently in its final design phase. The study presented here deals with design options and Remote Handling issues related to the waveguide assembly, an ensemble of mm-wave transmission line components mounted on a vacuum flange. This flange is part of the primary vacuum boundary of the ITER vessel. This paper describes the preliminary assessment of the RH compatibility of the sub-assembly, and a conceptual description of the maintenance actions to be performed on it. A comparison between two possible configurations for the tapers is reported: a waveguide integrated design, in which tapers are integral part of the in-plug waveguide, and an auxiliary shield integrated design, in which the tapers are integrated into the Auxiliary Shield. An important aspect in the design is to ensure the Remote Handling compatibility. Due to lack of space and limited dexterity of the slave manipulator, the general approach in defining the maintenance strategy is to avoid breaking the interfaces of the different components of the assembly, and to extract it from the Upper Port Plug as a single entity. An assessment on possible replacement procedure and the required tools are presented here

  13. Influence of visual feedback on human task performance in ITER remote handling

    International Nuclear Information System (INIS)

    Highlights: ► The performance of human operators in an ITER-like test facility for remote handling. ► Different sources of visual feedback influence how fast one can complete a maintenance task. ► Insights learned could be used in design of operator work environment or training procedures. - Abstract: In ITER, maintenance operations will be largely performed by remote handling (RH). Before ITER can be put into operation, safety regulations and licensing authorities require proof of maintainability for critical components. Part of the proof will come from using standard components and procedures. Additional verification and validation is based on simulation and hardware tests in 1:1 scale mockups. The Master Slave manipulator system (MS2) Benchmark Product was designed to implement a reference set of maintenance tasks representative for ITER remote handling. Experiments were performed with two versions of the Benchmark Product. In both experiments, the quality of visual feedback varied by exchanging direct view with indirect view (using video cameras) in order to measure and analyze its impact on human task performance. The first experiment showed that both experienced and novice RH operators perform a simple task significantly better with direct visual feedback than with camera feedback. A more complex task showed a large variation in results and could not be completed by many novice operators. Experienced operators commented on both the mechanical design and visual feedback. In a second experiment, a more elaborate task was tested on an improved Benchmark product. Again, the task was 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.

  14. The remote handling compatibility analysis of the ITER generic upper port plug structure

    International Nuclear Information System (INIS)

    Highlights: • We describe the remote handling compatibility of the ITER generic upper port plug. • Concepts are presented of specific design solutions to improve RH compatibility. • Simulation in VR of the GUPP DSM replacement indicates possible collisions. • Specific tooling concepts are proposed for GUPP handling equipment for the hot cell. - Abstract: The ITER diagnostics generic upper port plug (GUPP) is developed as a standardized design for all diagnostic upper port plugs, in which a variety of payloads can be mounted. Here, the remote handling compatibility analysis (RHCA) of the GUPP design is presented that was performed for the GUPP final design review. The analysis focuses mainly on the insertion and extraction procedure of the diagnostic shield module (DSM), a removable cassette that contains the diagnostic in-vessel components. It is foreseen that the DSM is a replaceable component – the procedure of which is to be performed inside the ITER hot cell facility (HCF), where the GUPP can be oriented in a vertical position. The DSM removal procedure in the HCF consists of removing locking pins, an M30 sized shoulder bolt and two electrical straps through the use of a dexterous manipulator, after which the DSM is lifted out of the GUPP by an overhead crane. For optimum access to its internals, the DSM is mounted in a handling device. The insertion of a new or refurbished DSM follows the reverse procedure. The RHCA shows that the GUPP design requires a moderate amount of changes to become fully compatible with RH maintenance requirements

  15. Preliminary Project Execution Plan for the Remote-Handled Low-Level Waste Disposal Project

    International Nuclear Information System (INIS)

    This preliminary project execution plan (PEP) defines U.S. Department of Energy (DOE) project objectives, roles and responsibilities of project participants, project organization, and controls to effectively manage acquisition of capital funds for construction of a proposed remote-handled low-level waste (LLW) disposal facility at the Idaho National Laboratory (INL). The plan addresses the policies, requirements, and critical decision (CD) responsibilities identified in DOE Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets.' This plan is intended to be a 'living document' that will be periodically updated as the project progresses through the CD process to construction and turnover for operation.

  16. Development of a remote handling system for replacement of armor tiles in the Fusion Experimental Reactor

    International Nuclear Information System (INIS)

    The armor tiles of the Fusion Experimental Reactor (FER) planned by JAERI are categorized as scheduled maintenance components, since they are damaged by severe heat and particle loads from the plasma during operation. A remote handling system is thus required to replace a large number of tiles rapidly in the highly activated reactor. However, the simple teaching-playback method cannot be adapted to this system because of deflection of the tiles caused by thermal deformation and so on. We have developed a control system using visual feedback control to adapt to this deflection and an end-effector for a single arm. We confirm their performance in tests. (orig.)

  17. Joint Working Group-39, Manufacturing Technology Subworking Group-F, remote handling and automation

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, R.D.

    1995-02-01

    The terms of reference were reviewed and continue to encompass the scope of activities of the SUBWOG. No revisions to the terms of reference were proposed. The list of site contacts who should receive copies of SUBWOG correspondence and meeting minutes was reviewed and updated. Documents exchanged related to the meeting include: Minutes of the sixth SUBOG 39F meeting; transactions of the fifth topical meeting on robotics and remote handling; data on manipulators was forwarded to LLNL from the robotics group at AEA Harwell; and the specifications of the duct remediation robot from the Rocky Flats Plant.

  18. Demonstration of a remotely operated TRU waste size-reduction and material handling process

    International Nuclear Information System (INIS)

    Noncombustible 238Pu and 239Pu waste is generated as a result of normal operation and decommissioning activity at the Savannah River Plant and is being retrievably stored at the Plant. As part of the long-term plant to process the wastes for permanent disposal, a remote size-reduction and material handling process is being cold-tested at Savannah River Laboratory to provide design support for the Plant TRU Waste Facility scheduled to be completed in 1992. Testing of the equipment in a complete cold demonstration, known as the Components Test Facility, is scheduled to begin in November 1986 and will run through 1987

  19. Concept design of divertor remote handling system for the FAST machine

    Energy Technology Data Exchange (ETDEWEB)

    Di Gironimo, G., E-mail: giuseppe.digironimo@unina.it [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, 80125 Napoli (Italy); Labate, C.; Renno, F. [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, 80125 Napoli (Italy); Brolatti, G.; Crescenzi, F.; Crisanti, F. [CR ENEA Frascati, Via E. Fermi 27, Frascati (RM) (Italy); Lanzotti, A. [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, 80125 Napoli (Italy); Lucca, F. [LT Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Siuko, M. [VTT Systems Engineering, Tekniikankatu 1, 33720 Tampere (Finland)

    2013-10-15

    The paper presents a concept design of a remote handling (RH) system oriented to maintenance operations on the divertor second cassette in FAST, a satellite of ITER tokamak. Starting from ITER configuration, a suitably scaled system, composed by a cassette multifunctional mover (CMM) connected to a second cassette end-effector (SCEE), can represent a very efficient solution for FAST machine. The presence of a further system able to open the divertor port, used for RH aims, and remove the first cassette, already aligned with the radial direction of the port, is presumed. Although an ITER-like system maintains essentially shape and proportions of its reference configuration, an appropriate arrangement with FAST environment is needed, taking into account new requirements due to different dimensions, weights and geometries. The use of virtual prototyping and the possibility to involve a great number of persons, not only mechanical designers but also physicist, plasma experts and personnel assigned to remote handling operations, made them to share the multiphysics design experience, according to a concurrent engineering approach. Nevertheless, according to the main objective of any satellite tokamak, such an approach benefits the study of enhancements to ITER RH system and the exploration of alternative solutions.

  20. TechnoFusion, a relevant facility for fusion technologies: The remote handling area

    International Nuclear Information System (INIS)

    The future commercial production of electricity based on thermonuclear fusion requires the development of a number of research projects in addition to ITER, mostly related to the development of long-term technologies needed for future fusion reactors. Among the priority areas identified in the framework of international fusion research programmes, it is projected to build a specific research centre for fusion technologies in Spain. The so-called National Centre for Fusion Technologies, TechnoFusion, will be equipped with a large number of top facilities for the fusion technological development. The research activities will be focused primarily on several areas, including the implementation of advanced manufacturing technologies, the evaluation of radiation effects on low activation structural and functional materials, the in-beam and out-beam characterization of materials, the development of robotics and automated systems for remote handling, studies of liquid metal technologies and computer simulation. Currently no similar facilities to TechnoFusion exist, so it will provide more realistic tests than those available in other facilities up today, helping in the fast track to DEMO and IFMIF. The present document summarizes a review of the different facilities to be included in TechnoFusion, with special emphasis on those proposed for remote handling (RH) applications. The paper will review the technical specifications concerning the RH facility, the analysis of the mock-up components and tests to be performed, and the relevance of the RH lab capabilities, particularly the required equipment under irradiation conditions.

  1. Preliminary concept design of the divertor remote handling system for DEMO power plant

    Energy Technology Data Exchange (ETDEWEB)

    Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Di Gironimo, G. [ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

    2014-11-15

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.

  2. Preliminary concept design of the divertor remote handling system for DEMO power plant

    International Nuclear Information System (INIS)

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations

  3. Progress in the design, R and D and procurement preparation of the ITER Divertor Remote Handling System

    International Nuclear Information System (INIS)

    Highlights: •The ITER Divertor Remote Handling System (DRHS) reference design is presented. •Different R and D activities that have contributed to the development and validation of the current reference design are reported. •The DRHS turns to be a unique system in terms of complexity due to size of the to-be-handled components, the novelty of the remote operations and the operational conditions. -- Abstract: The ITER Divertor Remote Handling System (DRHS) consists of a number of dedicated remote handling equipment and tooling that will provide the means to perform the exchange of the divertor system in a full-remote way. In order to achieve this objective the DRHS will need to perform a number of novel and complex remote operations in a contaminated and space-constrained environment, in rather poor lightening conditions. Fusion for Energy has recently launched the tendering phase for the in-kind procurement of the DRHS. The procurement is based on a set of system requirements and functional specifications supported by a reference design which are presented and discussed in this paper along with the main outcomes of the different R and D activities that have contributed to the development and validation of the current reference design

  4. Progress in the design, R and D and procurement preparation of the ITER Divertor Remote Handling System

    Energy Technology Data Exchange (ETDEWEB)

    Esqué, Salvador, E-mail: Salvador.Esque@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Hille, Carine van; Ranz, Roberto; Damiani, Carlo [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Palmer, Jim; Hamilton, David [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France)

    2014-10-15

    Highlights: •The ITER Divertor Remote Handling System (DRHS) reference design is presented. •Different R and D activities that have contributed to the development and validation of the current reference design are reported. •The DRHS turns to be a unique system in terms of complexity due to size of the to-be-handled components, the novelty of the remote operations and the operational conditions. -- Abstract: The ITER Divertor Remote Handling System (DRHS) consists of a number of dedicated remote handling equipment and tooling that will provide the means to perform the exchange of the divertor system in a full-remote way. In order to achieve this objective the DRHS will need to perform a number of novel and complex remote operations in a contaminated and space-constrained environment, in rather poor lightening conditions. Fusion for Energy has recently launched the tendering phase for the in-kind procurement of the DRHS. The procurement is based on a set of system requirements and functional specifications supported by a reference design which are presented and discussed in this paper along with the main outcomes of the different R and D activities that have contributed to the development and validation of the current reference design.

  5. Processing Plan for Potentially Reactive/Ignitable Remote Handled Transuranic Waste at the Idaho Cleanup Project - 12090

    International Nuclear Information System (INIS)

    Remote Handle Transuranic (RH-TRU) Waste generated at Argonne National Laboratory - East, from the examination of irradiated and un-irradiated fuel pins and other reactor materials requires a detailed processing plan to ensure reactive/ignitable material is absent to meet WIPP Waste Acceptance Criteria prior to shipping and disposal. The Idaho Cleanup Project (ICP) approach to repackaging Lot 2 waste and how we ensure prohibited materials are not present in waste intended for disposal at Waste Isolation Pilot Plant 'WIPP' uses an Argon Repackaging Station (ARS), which provides an inert gas blanket. Opening of the Lot 2 containers under an argon gas blanket is proposed to be completed in the ARS. The ARS is an interim transition repackaging station that provides a mitigation technique to reduce the chances of a reoccurrence of a thermal event prior to rendering the waste 'Safe'. The consequences, should another thermal event be encountered, (which is likely) is to package the waste, apply the reactive and or ignitable codes to the container, and store until the future treatment permit and process are available. This is the same disposition that the two earlier containers in the 'Thermal Events' were assigned. By performing the initial handling under an inert gas blanket, the waste can sorted and segregate the fines and add the Met-L-X to minimize risk before it is exposed to air. The 1-gal cans that are inside the ANL-E canister will be removed and each can is moved to the ARS for repackaging. In the ARS, the 1-gal can is opened in the inerted environment. The contained waste is sorted, weighed, and visually examined for non compliant items such as unvented aerosol cans and liquids. The contents of the paint cans are transferred into a sieve and manipulated to allow the fines, if any, to be separated into the tray below. The fines are weighed and then blended with a minimum 5:1 mix of Met-L-X. Other debris materials found are segregated from the cans into containers

  6. Apparatus for remote handling of materials. [mixing or analyzing dangerous chemicals

    Science.gov (United States)

    Kimball, R. B.; Hodder, D. T.; Wrinkle, W. W. (Inventor)

    1974-01-01

    Apparatus for remote handling of materials are described. A closed housing is provided with first and second containers and first and second reservoirs for holding materials to be mixed. The materials are transferable from the reservoirs to the first container where they are mixed. The mixed materials are then conveyed from the first container to the second container preferably by dumping the mixed materials into a funnel positioned over the second container. The second container is then moved to a second position for analysis of the mixed materials. For example, the materials may be ignited and the flame analyzed. Access, such as a sight port, is provided in the housing at the analysis position. The device provides a simple and inexpensive apparatus for safely mixing a pyrophoric material and an oxidizer which together form a thermite type mixture that burns to produce a large quantity of heat and light.

  7. Automation of remote handling in uranium and mixed oxide fuel element fabrication plants

    International Nuclear Information System (INIS)

    The subject of the analyses are plants for the fabrication or uranium oxide and uranium-plutonium mixed oxide fuel elements. The reference basis of the paper is an overview of the state-of-the-art of manufacturing technologies with regard to automation and remote handling during fuel element fabrication in national and foreign plants, and in comparabel sectors of conventional technologies. Proceeding from ambient dose rates, residence times, and technical conditions or individual doses at typical work-places during fuel element fabrication, work processes are pointed out which, taking into account technical possibilities, should be given priority when automating, and technical solutions for it are sought. Advantages and disadvantages of such measures are outlined, and reduction of radiation exposure is shown (example: mixed oxide fuel fabrication plant at Hanau). (orig./HP)

  8. Preliminary Project Execution Plan for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2011-05-01

    This preliminary project execution plan (PEP) defines U.S. Department of Energy (DOE) project objectives, roles and responsibilities of project participants, project organization, and controls to effectively manage acquisition of capital funds for construction of a proposed remote-handled low-level waste (LLW) disposal facility at the Idaho National Laboratory (INL). The plan addresses the policies, requirements, and critical decision (CD) responsibilities identified in DOE Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets.' This plan is intended to be a 'living document' that will be periodically updated as the project progresses through the CD process to construction and turnover for operation.

  9. Calibration and compensation of deflections and compliances in remote handling equipment configurations

    International Nuclear Information System (INIS)

    This paper presents a generic method of calibrating and compensating remote handling system configurations subject to manufacturing and assembly tolerances, deflections and compliances. A method consists of kinematic part and non-kinematic part. A kinematic calibration algorithm is presented for finding the values of kinematic model errors by measuring the end-effector Cartesian position. This is a conventional way to calibrate industrial robots. However, in this case the kinematic calibration is not able to compensate flaws fully due to large deflections and compliances caused by a massive Cassette payload (approx. 9 ton). Positioning error at the furthest point of the cassette before any compensation was 80 mm. Therefore, extra compensation must be introduced in addition to a kinematic calibration. A kinematic calibration together with an extra compensation is a demanding task to carry out. The resulting complex compensation function has to be such that it can be implemented in real-time Cassette Multifunctional Mover (CMM) control system software.

  10. Interoperability of remote handling control system software modules at Divertor Test Platform 2 using middleware

    Energy Technology Data Exchange (ETDEWEB)

    Tuominen, Janne, E-mail: janne.m.tuominen@tut.fi [Tampere University of Technology, Department of Intelligent Hydraulics and Automation, Tampere (Finland); Rasi, Teemu; Mattila, Jouni [Tampere University of Technology, Department of Intelligent Hydraulics and Automation, Tampere (Finland); Siuko, Mikko [VTT, Technical Research Centre of Finland, Tampere (Finland); Esque, Salvador [F4E, Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla2, 08019, Barcelona (Spain); Hamilton, David [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► The prototype DTP2 remote handling control system is a heterogeneous collection of subsystems, each realizing a functional area of responsibility. ► Middleware provides well-known, reusable solutions to problems, such as heterogeneity, interoperability, security and dependability. ► A middleware solution was selected and integrated with the DTP2 RH control system. The middleware was successfully used to integrate all relevant subsystems and functionality was demonstrated. -- Abstract: This paper focuses on the inter-subsystem communication channels in a prototype distributed remote handling control system at Divertor Test Platform 2 (DTP2). The subsystems are responsible for specific tasks and, over the years, their development has been carried out using various platforms and programming languages. The communication channels between subsystems have different priorities, e.g. very high messaging rate and deterministic timing or high reliability in terms of individual messages. Generally, a control system's communication infrastructure should provide interoperability, scalability, performance and maintainability. An attractive approach to accomplish this is to use a standardized and proven middleware implementation. The selection of a middleware can have a major cost impact in future integration efforts. In this paper we present development done at DTP2 using the Object Management Group's (OMG) standard specification for Data Distribution Service (DDS) for ensuring communications interoperability. DDS has gained a stable foothold especially in the military field. It lacks a centralized broker, thereby avoiding a single-point-of-failure. It also includes an extensive set of Quality of Service (QoS) policies. The standard defines a platform- and programming language independent model and an interoperability wire protocol that enables DDS vendor interoperability, allowing software developers to avoid vendor lock-in situations.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sibois, Romain, E-mail: romain.sibois@vtt.fi [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Määttä, Timo; Siuko, Mikko [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Mattila, Jouni [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland)

    2014-10-15

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

  12. Interoperability of remote handling control system software modules at Divertor Test Platform 2 using middleware

    International Nuclear Information System (INIS)

    Highlights: ► The prototype DTP2 remote handling control system is a heterogeneous collection of subsystems, each realizing a functional area of responsibility. ► Middleware provides well-known, reusable solutions to problems, such as heterogeneity, interoperability, security and dependability. ► A middleware solution was selected and integrated with the DTP2 RH control system. The middleware was successfully used to integrate all relevant subsystems and functionality was demonstrated. -- Abstract: This paper focuses on the inter-subsystem communication channels in a prototype distributed remote handling control system at Divertor Test Platform 2 (DTP2). The subsystems are responsible for specific tasks and, over the years, their development has been carried out using various platforms and programming languages. The communication channels between subsystems have different priorities, e.g. very high messaging rate and deterministic timing or high reliability in terms of individual messages. Generally, a control system's communication infrastructure should provide interoperability, scalability, performance and maintainability. An attractive approach to accomplish this is to use a standardized and proven middleware implementation. The selection of a middleware can have a major cost impact in future integration efforts. In this paper we present development done at DTP2 using the Object Management Group's (OMG) standard specification for Data Distribution Service (DDS) for ensuring communications interoperability. DDS has gained a stable foothold especially in the military field. It lacks a centralized broker, thereby avoiding a single-point-of-failure. It also includes an extensive set of Quality of Service (QoS) policies. The standard defines a platform- and programming language independent model and an interoperability wire protocol that enables DDS vendor interoperability, allowing software developers to avoid vendor lock-in situations

  13. Safety Design Strategy for the Remote Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Chirstensen

    2012-04-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3A, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3A and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Remote-Handled Low-Level Waste Disposal Project.

  14. Safety Design Strategy for the Remote Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Chirstensen

    2012-08-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3A, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3A and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Remote-Handled Low-Level Waste Disposal Project.

  15. Safety Design Strategy for the Remote Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Chirstensen

    2015-03-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3A, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3A and DOE Order 420.1C, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Remote-Handled Low-Level Waste Disposal Project.

  16. Safety Design Strategy for the Remote Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Gary Mecham

    2009-10-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3A, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3A and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Remote-Handled Low-Level Waste Disposal Project.

  17. Divertor cassette locking system remote handling trials with WHMAN at DTP2

    Energy Technology Data Exchange (ETDEWEB)

    Lyytikäinen, Ville; Kinnunen, Pasi; Koivumäki, Janne; Mattila, Jouni [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Siuko, Mikko [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Esque, Salvador [F4E, Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla2, 08019, Barcelona (Spain); Palmer, Jim, E-mail: ville.lyytikainen@tut.fi [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► RH requirements were developed from operator feedback, potential problem analysis and task description. ► Tools were designed according to these RH specific requirements. ► Two RH capable were developed and their functionality was verified at DPT2. -- Abstract: A key ITER maintenance activity is the exchange of the divertor cassettes. The current major step in this programme involves the full scale physical test facility, namely divertor test platform 2 (DTP2), in Tampere, Finland. The objective of the DTP2 is the design and proof of concept studies of various remote handling (RH) device prototypes and their RH control systems, but is also important to define principles for standardizing control systems and methods around the ITER maintenance equipment. The development process of divertor cassette locking system (CLS) RH Tool prototypes is presented in this paper. The validation of the developed CLS Tool prototypes is accomplished in RH trials at DTP2. For this RH Trial, a CLS task description (TD) and tool prototypes were developed, manufactured and, finally, tested under remote operations. These tools, designed to be operated by water hydraulic manipulator (WHMAN), are water hydraulic jack (WHJ), pin tool (PT) and wrench tool (WT)

  18. Results from simulated remote-handled transuranic waste experiments at the Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

    Multi-year, simulated remote-handled transuranic waste (RH TRU, nonradioactive) experiments are being conducted underground in the Waste Isolation Pilot-Plant (WIPP) facility. These experiments involve the near-reference (thermal and geometrical) testing of eight full size RH TRU test containers emplaced into horizontal, unlined rock salt boreholes. Half of the test emplacements are partially filled with bentonite/silica-sand backfill material. All test containers were electrically heated at about 115 W/each for three years, then raised to about 300 W/each for the remaining time. Each test borehole was instrumented with a selection of remote-reading thermocouples, pressure gages, borehole vertical-closure gages, and vertical and horizontal borehole-diameter closure gages. Each test emplacements was also periodically opened for visual inspections of brine intrusions and any interactions with waste package materials, materials sampling, manual closure measurements, and observations of borehole changes. Effects of heat on borehole closure rates and near-field materials (metals, backfill, rock salt, and intruding brine) interactions were closely monitored as a function of time. This paper summarizes results for the first five years of in situ test operation with supporting instrumentation and laboratory data and interpretations. Some details of RH TRU waste package materials, designs, and assorted underground test observations are also discussed. Based on the results, the tested RH TRU waste packages, materials, and emplacement geometry in unlined salt boreholes appear to be quite adequate for initial WIPP repository-phase operations

  19. Failure mode and effect analysis for remote handling transfer systems of ITER FE

    International Nuclear Information System (INIS)

    A Failure Mode and Effect Analysis (FMEA) at component level was done to study safety relevant implications arising from possible failures in performing Remote Handling (RH) operations. Autonomous air cushion transporter, pallet, sealed casks and tractor movers needed for port plug mounting/dismantling operation were analysed. For each sub-system, the breakdown of significant components was outlined and, for each component, possible failure modes have been investigated pointing out possible causes, possible actions to prevent the causes, consequences and actions to prevent or mitigate consequences. Off-normal events which may result in hazardous consequences for the public and the environment have been defined as Postulated Initiating Events (PIEs). Two safety-relevant PIEs have been defined by assessing elementary failures related to the analysed system. Each PIE has been discussed in order to qualitatively identify accident sequences arising from each of them. The two PIEs are: - RHP Radioactive products (fraction of Dust and T implanted in VV) into Port Cell during RH operations for breach in ''VV+cask'' isolating boundary. - RHG Cask stop and radioactive products (fraction of Dust and T implanted in VV) release into Gallery due to Cask leakage during transportation to Hot Cell. At first glance the consequences of such accidents in terms of radioactive releases should be within the assessment of consequences performed for other studies. Nevertheless, further deterministic analysis could be required to determine response of safety systems (e.g.: efficiency of ventilation systems, isolation of HVAC) and effectiveness of rescue operations in mitigating the consequences and risks for workers. Precisely, even if the two PIEs do not lead to significant radioactive release to the environment, spreading of contamination inside the building and the operating areas can be induced. Consequently, for maintenance and/or decontamination activities, over radiation exposure to

  20. National Environmental Policy Act Compliance Strategy for the Remote-Handled Low-level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Peggy Hinman

    2010-10-01

    The U.S. Department of Energy (DOE) needs to have disposal capability for remote-handled low level waste (LLW) generated at the Idaho National Laboratory (INL) at the time the existing disposal facility is full or must be closed in preparation for final remediation of the INL Subsurface Disposal Area in approximately the year 2017.

  1. Remote Handled WIPP Canisters at Los Alamos National Laboratory Characterized for Retrieval

    International Nuclear Information System (INIS)

    The Los Alamos National Laboratory (LANL) is pursuing retrieval, transportation, and disposal of 16 remote handled transuranic waste canisters stored below ground in shafts since 1994. These canisters were retrievably stored in the shafts to await Nuclear Regulatory Commission certification of the Model Number RH-TRU 72B transportation cask and authorization of the Waste Isolation Pilot Plant (WIPP) to accept the canisters for disposal. Retrieval planning included radiological characterization and visual inspection of the canisters to confirm historical records, verify container integrity, determine proper personnel protection for the retrieval operations, provide radiological dose and exposure rate data for retrieval operations, and to provide exterior radiological contamination data. The radiological characterization and visual inspection of the canisters was performed in May 2006. The effort required the development of remote techniques and equipment due to the potential for personnel exposure to radiological doses approaching 300 R/hr. Innovations included the use of two nested 1.5 meter (m) (5-feet [ft]) long concrete culvert pipes (1.1-m [42 inch (in.)] and 1.5-m [60-in] diameter, respectively) as radiological shielding and collapsible electrostatic dusting wands to collect radiological swipe samples from the annular space between the canister and shaft wall. Visual inspection indicated that the canisters are in good condition with little or no rust, the welded seams are intact, and ten of the canisters include hydrogen gas sampling equipment on the pintle that will have to be removed prior to retrieval. The visual inspection also provided six canister identification numbers that matched historical storage records. The exterior radiological data indicated alpha and beta contamination below LANL release criteria and radiological dose and exposure rates lower than expected based upon historical data and modeling of the canister contents. (authors)

  2. Irradiation tests of critical components for remote handling system in gamma radiation environment

    International Nuclear Information System (INIS)

    This report covers the gamma ray irradiation tests according to the Agreement of ITER R and D Task (T35) in 1994 and describes radiation hardness of the standard components for the ITER remote handling system which are categorized into the robotics (Subtask-1), the viewing system (Subtask-2) and the common components (Subtask-3). The gamma ray irradiation tests have been conducted using No.2 and No.3 cells at the cobalt building of Takasaki Establishment in JAERI. The radiation source is cobalt sixty (Co-60), and the maximum dose rate of No.2 and No.3 cells is about 1x106 R/h and 2x106 R/h, respectively. The environmental conditions of the irradiation tests are described below and all of components excepting electrical wires have been tested in the No.2 cell. [No.2 cell : Atmosphere and ambient temperature No.3 cell : Nitrogen gas and 250degC] As a whole, many of components have been irradiated up to the rated dose of around 1x1010 rads and the following main results are obtained. The developed AC servo motor and periscope for radiation use have shown excellent durability with the radiation hardness tolerable for more than 109 rads. An electrical connector compatible with remote operation has also shown no degradation of electrical characteristics after the irradiation of 1010 rads. As for polyimide insulated wires, the mechanical and electrical characteristics are not degradated after the irradiation of 109 rads and more radiation hardness can be expected than the anticipation. On the contrary, standard position sensors such as rotary encoder show extremely low radiation hardness and further efforts have to be made for improvements. (J.P.N.)

  3. Cultural Resource Investigations for the Remote Handled Low Level Waste Facility at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Brenda R. Pace; Hollie Gilbert; Julie Braun Williams; Clayton Marler; Dino Lowrey; Cameron Brizzee

    2010-06-01

    The U. S. Department of Energy, Idaho Operations Office is considering options for construction of a facility for disposal of Idaho National Laboratory (INL) generated remote-handled low-level waste. Initial screening has resulted in the identification of two recommended alternative locations for this new facility: one near the Advanced Test Reactor (ATR) Complex and one near the Idaho Comprehensive Environmental Response, Compensation, and Liability Act Disposal Facility (ICDF). In April and May of 2010, the INL Cultural Resource Management Office conducted archival searches, intensive archaeological field surveys, and initial coordination with the Shoshone-Bannock Tribes to identify cultural resources that may be adversely affected by new construction within either one of these candidate locations. This investigation showed that construction within the location near the ATR Complex may impact one historic homestead and several historic canals and ditches that are potentially eligible for nomination to the National Register of Historic Places. No resources judged to be of National Register significance were identified in the candidate location near the ICDF. Generalized tribal concerns regarding protection of natural resources were also documented in both locations. This report outlines recommendations for protective measures to help ensure that the impacts of construction on the identified resources are not adverse.

  4. Cultural Resource Investigations for the Remote Handled Low Level Waste Facility at the Idaho National Laboratory

    International Nuclear Information System (INIS)

    The U. S. Department of Energy, Idaho Operations Office is considering options for construction of a facility for disposal of Idaho National Laboratory (INL) generated remote-handled low-level waste. Initial screening has resulted in the identification of two recommended alternative locations for this new facility: one near the Advanced Test Reactor (ATR) Complex and one near the Idaho Comprehensive Environmental Response, Compensation, and Liability Act Disposal Facility (ICDF). In April and May of 2010, the INL Cultural Resource Management Office conducted archival searches, intensive archaeological field surveys, and initial coordination with the Shoshone-Bannock Tribes to identify cultural resources that may be adversely affected by new construction within either one of these candidate locations. This investigation showed that construction within the location near the ATR Complex may impact one historic homestead and several historic canals and ditches that are potentially eligible for nomination to the National Register of Historic Places. No resources judged to be of National Register significance were identified in the candidate location near the ICDF. Generalized tribal concerns regarding protection of natural resources were also documented in both locations. This report outlines recommendations for protective measures to help ensure that the impacts of construction on the identified resources are not adverse.

  5. Performance Assessment for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Annette L. Schafer; A. Jeffrey Sondrup; Arthur S. Rood

    2012-05-01

    This performance assessment for the Remote-Handled Low-Level Radioactive Waste Disposal Facility at the Idaho National Laboratory documents the projected radiological dose impacts associated with the disposal of low-level radioactive waste at the facility. This assessment evaluates compliance with the applicable radiological criteria of the U.S. Department of Energy and the U.S. Environmental Protection Agency for protection of the public and the environment. The calculations involve modeling transport of radionuclides from buried waste to surface soil and subsurface media, and eventually to members of the public via air, groundwater, and food chain pathways. Projections of doses are calculated for both offsite receptors and individuals who inadvertently intrude into the waste after site closure. The results of the calculations are used to evaluate the future performance of the low-level radioactive waste disposal facility and to provide input for establishment of waste acceptance criteria. In addition, one-factor-at-a-time, Monte Carlo, and rank correlation analyses are included for sensitivity and uncertainty analysis. The comparison of the performance assessment results to the applicable performance objectives provides reasonable expectation that the performance objectives will be met

  6. Real-time markerless Augmented Reality for Remote Handling system in bad viewing conditions

    International Nuclear Information System (INIS)

    Remote Handling (RH) in harsh environments usually has to tackle the lack of sufficient visual feedback for the human operator due to the limited number of on-site cameras, the not optimized position of the cameras, the poor viewing angles, occlusion, failure, etc. Augmented Reality (AR) enables the user to perceive virtual computer-generated objects in a real scene. The most common goals usually include visibility enhancement and provision of extra information, such as positional data of various objects. The proposed AR system first recognizes and locates the markerless object by using a template based matching algorithm, and then augments the virtual model on top of the recognized item. The tracking algorithm is exploited for locating the object in a continuous sequence of frames. Conceptually, the template is found by computing the similarity between the template and the image frame, for all the relevant template poses (rotation and translation). As a case study, AR interface was displaying measured orientation and transformation of the Water Hydraulic Manipulator (WHMAN) Divertor preloading tool, in near real-time tracking. The bad viewing condition implies on the case when the view angle is such that the interesting features of the object are not in the field of view. The method in this paper was validated in concrete operational context at DTP2. The developed method proved to deliver robust positional and orientation information while augmenting and tracking the moving tool object.

  7. Command and Control application framework for interoperable heterogeneous ITER Remote Handling devices

    International Nuclear Information System (INIS)

    The ITER divertor Remote Handling maintenance concepts are under investigation at a full scale mockup facility, the Divertor Test Platform, located in Tampere, Finland. The key devices to carry out these operations are the (CMM), Cassette Toroidal Mover, Multi Axis Manipulators (MAM) and related tooling. Despite their differences, in essence all of these devices are manipulators with varying specifications. Hence, from the operator's point of view, they perform similar operations, such as following predefined position trajectories and lifting and carrying tools and equipment. This paper explores how to implement a software framework for a Command and Control application, otherwise known as Human Machine Interface (HMI), in such a way that it could be used to control different manipulators. A framework is presented that supports the combining of small, independent modules into a single cohesive HMI. It is then shown that a HMI can be built by combining modules that are common to these manipulators and modules that are specific to a single manipulator. This will not only minimize the programming effort but also reduce the training required of operators. To elaborate on the concept, an example HMI is implemented to control virtual models of CMM and MAM.

  8. Storage tube with a base adjustable for height by remote handling, particularly for spent fuel storage in pools

    International Nuclear Information System (INIS)

    One aim of this invention is the fabrication of a storage tube with a base adjustable for height by remote handling, for the in-pools storing of irradiated nuclear fuels. This device possesses the following features with respect to the mechanism for placing the base in the supporting position: - use of rotation without rectilinear friction or gears, - impossibility for dust to accumulate on the mechanism, - possible control by handling pole, - simplicity and low mass production cost. Such features can of course be used to advantage for the tubes to store elements of various lengths irrespective of the nuclear energy

  9. Overview of Bore Tools Systems for divertor remote maintenance of ITER

    International Nuclear Information System (INIS)

    Because of the radiation levels preventing direct, hands-on access to the machine components, maintenance work on ITER will eventually require the use of Remote Handling techniques. In particular, the replacement of components such as divertor and blanket modules will require the use of remote cutting, welding and Non Destructive Testing of water cooling pipes

  10. The Design and Development of Divertor Remote Handling Equipment for ITER

    International Nuclear Information System (INIS)

    A key ITER maintenance activity is the complete exchange of the divertor system at scheduled intervals, typically after every 3-4 years of plasma operations. In view of this, ITER divertor maintenance is classified as an RH Class 1 activity and as such, detailed design of the associated RH equipment and verification of its operation before ITER construction by way of prototypes and mock-ups, is considered an essential activity. Throughout the course of the ITER design activities one of the major focuses of the EU contribution has been the study and development of remote handling equipment (RHE) necessary for divertor exchange. This suite of RHE will include a number of heavy in-vessel robotic transporters (known as '' ssette movers ''), ex-vessel transfer casks and several general purpose dextrous manipulators used to deploy and operate task-specific RH tooling. The current major step in the divertor RH development programme for ITER involves the construction of a full scale physical test facility in which to demonstrate and refine RHE designs through the operation of prototypes closely replicating those proposed for ITER. This facility, designated the '' Divertor Test Platform 2 (DTP2) '', will be constructed in Tampere, Finland and operated by the Finnish Fusion Association, TEKES. Four separate procurement contracts are currently being executed within European Industry for the supply of the major DTP2 sub-systems namely, a mock-up of the ITER divertor region, a mock-up cassette, a prototype cassette mover and the mover control hardware. The control system software development is being carried out in parallel by staff of the DTP2 host organisation. These DTP2 sub-systems will be brought together in Tampere during the Autumn 2006 / Spring 2007 and the system is expected to be ready for RH trials on the second cassette handling process in the summer of that year. Measures to extend the facility in subsequent years to allow more extensive trials in a 30 degree

  11. Tritium contamination and decontamination study on materials for ITER remote handling equipment

    International Nuclear Information System (INIS)

    Several materials, lenses, dry bearings and cables were exposed to a tritiated moisture environment to study the behavior of tritium contamination on candidate materials for ITER remote handling equipment. To optimize the tritium removal procedure, decontamination experiments using a gas purge with three different moisture concentrations were also performed. The surface tritium concentrations of CeO2 containing alkaline barium glass (NB), CeO2 containing lead glass (LX) and synthetic quartz (Quartz) after the exposure experiments were 7.80, 10.94 and 0.67 Bq/cm2, respectively. It was found that the tritium concentration was influenced by the compositions of the materials. The concentrations of tritium on type 831 (solid lubrication material: graphite) and type 237 (solid lubrication material: tungsten disulfate) dry bearings after the exposure experiments were 89.80 and 31.78 Bq/cm2, respectively. The tritium concentration in an electric cable tested was 5014 Bq/g after HTO moisture exposure. The tritium concentrations of lenses, LX, as typical experimental results, decreased to 2.72, 4.42 and 3.89 Bq/cm2 by purging with the moist air, dry air and dry N2, respectively. The tritium concentrations of dry bearing, type 831 dropped to 6.61, 9.42 and 10.16 Bq/cm2 by the same three decontamination treatments, respectively. A large decontamination factor of 13.6 was achieved in the case of type 831 dry bearing with a moist air purge. The tritium concentration in the electric cable was 3236 Bq/g after a moist air purge, and the decontamination factor was as low as 1.6. Therefore, decontamination with a moist air purge is not so effective for the electric cable

  12. Logistics management for storing multiple cask plug and remote handling systems in ITER

    International Nuclear Information System (INIS)

    Highlights: ► We model the logistics management problem in ITER, taking into account casks of multiple typologies. ► We propose a method to determine the best position of the casks inside a given storage area. ► Our method obtains the sequence of operations required to retrieve or store an arbitrary cask, given its storage place. ► We illustrate our method with simulation results in an example scenario. -- Abstract: During operation, maintenance inside the reactor building at ITER (International Thermonuclear Experimental Reactor) has to be performed by remote handling, due to the presence of activated materials. Maintenance operations involve the transportation and storage of large, heavyweight casks from and to the tokamak building. The transportation is carried out by autonomous vehicles that lift and move beneath these casks. The storage of these casks face several challenges, since (1) the cask storage area is limited in space, and (2) all casks have to be accessible for transportation by the vehicles. In particular, casks in the storage area may block other casks, so that the former has to be moved to a temporary position to give way to the latter. This paper addresses the challenge of managing the logistics of cask storage, where casks may have different typologies. In particular, we propose an approach to (1) determine the best position of the casks inside the storage area, and to (2) obtain the sequence of operations required to retrieve and store an arbitrary cask from/to a given storage place. A combinatorial optimization approach is used to obtain solutions to both these problems. Simulation results illustrate the application of the proposed method to a simple scenario

  13. Mirrors for diagnostic and remote handling applications in ITER. Problems with commercial mirrors

    International Nuclear Information System (INIS)

    High quality mirrors for the optical UV - visible - NIR range will be required in ITER for diagnostic systems and remote handling applications. The commercially available high quality mirrors being considered for ITER applications consist of a thin evaporated aluminium layer on a solid glass substrate. To protect the delicate aluminium surface from damage the mirrors are covered (overcoated) with a controlled layer of transparent protective material (such as SiO and MgF2) of adequate thickness to obtain optimum optical constructive interference in a given wavelength range. Radiation enhanced degradation of reflectivity of different commercial mirrors as a function of irradiation temperature and time (dose), as well as enhanced corrosion in a humid environment are being studied. During this work problems have been encountered with the as-received commercial mirrors concerning reflectivity and overcoating. Mirrors provided as identical, in the same batch by the manufacturer, have very different reflectivity in the UV range, implying wide variation in the overcoating thickness and/or refractive index. Analysis of the surface material has shown varying amounts of SiO and SiO2 for nominally SiO overcoating. In another case the specified coating (magnesium fluoride) for an UV enhanced mirror was observed by XPS analysis to contain Hf, neither Mg nor F were found. It will therefore be very important to analyse composition and measure reflectivity for all commercial mirrors before use. The varying amount of SiO found in the overcoating helps to explain the very different rates of radiation enhanced corrosion of nominally identical mirrors, those with higher SiO content being more susceptible. To check this, reflectivity changes after gamma irradiation and corrosion effects have been measured for two types of mirrors made at CIEMAT, one with SiO2, and one with Al2O3 overcoating of Al on Pyrex glass. (author)

  14. Proximity measuring device with backscattering radiation usable noticeably in remote handling or robotics and related data processing system

    International Nuclear Information System (INIS)

    The invention is aimed at a proximity measuring device whose emitter, an electroluminescent diode, is controlled by control means to emit short duration ( 1A) flashes with periods higher than 100 microseconds. Emetter-object distance can be precisely measured on an 0-30 cm interval with the help of data processing of the response given by the proximity device receiver. This device can be used in remote handling and robotics

  15. Progress on the interface between UPP and CPRHS (Cask and Plug Remote Handling System) tractor/gripping tool for ITER

    International Nuclear Information System (INIS)

    Highlights: ► UPP interface requirements in the plug RH extraction/insertion for ITER. ► Analyze of maximum misalignment between port duct and port cell. ► Friction study between plug skids and VV port/ramp rails during the plug transfer. ► Definition of the tolerance in the plug skids to avoid the plug jamming. ► Concepts of gripping tools based on one gripping point and avoiding force feedback. -- Abstract: EFDA finances a training programme called Goal Oriented Training Programme for Remote Handling (GOT RH), whose goal is to train engineers in Remote Handling for ITER. As part of this training programme, the conceptual design of the mechanical interface between Upper Port Plug (UPP) and Cask and Plug Remote Handling System (CPRHS) as well as the conceptual design of the needed tools for UPP Remote Handling is carried out. The paper presents the conceptual design of the UPP/Gripping Tool Interface. This includes the conceptual design of the gripping tool for introducing/removing the UPP in/from the ITER port and the mechanical features on both sides of the UPP/Gripping Tool Interface (e.g. alignment features, mechanical connectors, fasteners). In order to develop the design of the interface between UPP and CPRHS it is necessary to first identify the functional requirements of the Transfer Cask System (TCS) and the CPRHS, such as required degrees of freedom (DoF), required performances of system, geometrical constraints, loading conditions, alignment requirements, RAMI requirements. These requirements are the input data for the design of the interface between UPP and gripping tool and some of them are also described in the paper

  16. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    A conceptual design has been carried out for an Engineering Test Station (ETS) which will provide structural support and utilities/instrumentation services for blanket modules positioned adjacent to the vacuum vessel of the TFTR (Tokamak Fusion Test Reactor). The ETS is supported independently from the Test Cell floor. The ETS module support platform is constructed of fiberglass to eliminate electromagnetic interaction with the pulsed tokamak fields. The ETS can hold blanket modules with dimensions up to 78 cm in width, 85 cm in height, and 105 cm in depth, and with a weight up to 4000 kg. Interfaces for all utility and instrumentation requirements are made via a shield plug in the TFTR igloo shielding. The modules are readily installed or removed by means of TFTR remote handling equipment

  17. The application of advanced remote systems technology to future waste handling facilities: Waste Systems Data and Development Program

    International Nuclear Information System (INIS)

    The Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) has been advancing the technology in remote handling and remote maintenance of in-cell systems planned for future US nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor are directly applicable to the in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The ORNL developments are based on the application of teleoperated force-reflecting servomanipulators controlled by an operator completely removed from the hazardous environment. These developments address the nonrepetitive nature of remote maintenance in the unstructured environments encountered in a waste handling facility. Employing technological advancements in dexterous manipulators, as well as basic design guidelines that have been developed for remotely maintained equipment and processes, can increase operation and maintenance system capabilities, thereby allowing the attainment of two FWMS major objectives: decreasing plant personnel radiation exposure and increasing plant availability by decreasing the mean-time-to-repair in-cell maintenance and process equipment. 5 refs., 7 figs

  18. 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.

  19. The application of mature dry storage technology and remote handling robotics to nuclear plant extension, clean-up and decommissioning

    International Nuclear Information System (INIS)

    This paper reviews a mature dry storage technology developed by GEC ALSTHOM Engineering Systems Limited (GAES) which offers a passive, economical and licensable method of providing irradiated fuel storage capacity at operational nuclear power stations. The evolution of the modular vault dry store (MVDS) technology has taken place over 25 years of operational experience, culminating in a product which meets all of the concerns of licensing authorities regarding safety and fuel integrity. The application of remote handling robotics to nuclear fuel and active component handling as a routine process rather than as an intervention technique is also reviewed. The growth of the application of this technology is governed by several factors which include: statutory requirements, safety assurance, risk reduction and economic pressures. The availability of a mature MVDS technology with an evolving process-capable robotics technology opens up opportunities for exploring proven UK products. (Author)

  20. Shipping Remote Handled Transuranic Waste to the Waste Isolation Pilot Plant - An Operational Experience

    International Nuclear Information System (INIS)

    On January 18, 2007, the first ever shipment of Remote Handled Transuranic (RH TRU) waste left the gate at the Idaho National Laboratory (INL), headed toward the Waste Isolation Pilot Plant (WIPP) for disposal, thus concluding one of the most stressful, yet rewarding, periods the authors have ever experienced. The race began in earnest on October 16, 2006, with signature of the New Mexico Environment Department Secretary's Final Order, ruling that the '..draft permit as changed is hereby approved in its entirety.' This established the effective date of the approved permit as November 16, 2006. The permit modification was a consolidation of several Class 3 modification requests, one of which included incorporation of RH TRU requirements and another of which incorporated the requirements of Section 311 of Public Law 108-137. The obvious goal was to complete the first shipment by November 17. While many had anticipated its approval, the time had finally come to actually implement, and time seemed to be the main item lacking. At that point, even the most aggressive schedule that could be seriously documented showed a first ship date in March 2007. Even though planning for this eventuality had started in May 2005 with the arrival of the current Idaho Cleanup Project (ICP) contractor (and even before that), there were many facility and system modifications to complete, startup authorizations to fulfill, and many regulatory audits and approvals to obtain before the first drum could be loaded. Through the dedicated efforts of the ICP workers, the partnership with Department of Energy (DOE) - Idaho, the coordinated integration with the Central Characterization Project (CCP), the flexibility and understanding of the regulatory community, and the added encouragement of DOE - Carlsbad Field Office and at Headquarters, the first RH TRU canister was loaded on December 22, 2006. Following final regulatory approval on January 17, 2007, the historic event finally occurred the

  1. Fast Flux Test Facility interim examination and maintenance cell contaminated sodium recovery system: Remote handling design consideration

    International Nuclear Information System (INIS)

    The Westinghouse Hanford Company is installing a remotely operated Contaminated Sodium Recovery System (CSRS) at the Fast Flux Test Facility (FFTF) located in Richland, Washington. The CSRS will recover activated sodium that accumulates in fuel transfer machines during core component transfer operations. Drip pots from the FFTF fuel handling machines will be delivered to the shielded, argon-inerted Interim Examination and Maintenance (IEM) Cell, a hot cell located in the FFTF containment structure. Installation of the CSRS replaces a previously manual operation that required disposal of radioactive sodium with a completely remote operation that will return sodium to service in the plant. The CSRS will minimize the accumulation of hazardous waste and reduce personnel exposure to radioactive materials. Equipment for the CSRS is currently being fabricated and tested before installation in the IEM Cell. 6 figs

  2. Overview of remote handling technologies developed for inspection and maintenance of spent fuel management facilities in France

    International Nuclear Information System (INIS)

    In the facilities of the end of the nuclear fuel cycle, like spent fuel storage pools, reprocessing plants, Plutonium-based fuel manufacturing plants or waste temporary storage units, materials handling must be carried out remotely, taking into account the nuclear radiating environment. In addition to the automation requirement, robotics equipment in the nuclear industry must be substituted to human operators in order to respect the ALARA principle. More over, remote handling technologies aim to improve the working conditions, as well as the quality of the work achieved by the operators. Ten years ago, COGEMA (AREVA Group) and CEA (French Atomic Energy Agency) started an ambitious R and D program in robotics and remote handling technologies applied to COGEMA spent fuel management facilities in France, with the aim to cover the requirements of the different plant life cycle steps. The paper gives an overview of the important developments that have been carried out by CEA and then transferred to the COGEMA industrial group. The range includes the next generation of servo-manipulators, long range inspection tools and carriers, nuclear versions of industrial robots, radiation hardened electronic systems, interactive environment modeling tools, as well as force-feedback master-slave generic control software for tele-operation systems. Some applications of this development are presented in the paper: - rad-hard electronic modules for robotic equipment which are used by COGEMA in high radiating environment; - long reach articulated carrier for inspection of spent full management blind cells; - new electrical force feedback master/slave system to improve the tele-operation of standard tele-manipulators; - generic control software for tele-manipulators. The results of the robotic program carried out by COGEMA and CEA have been very valuable for the introduction of new technologies inside nuclear industry. Innovative products and sub-systems can be integrated now in a large

  3. Selected solutions and design features from the design of remotely handled filters and the technology of remote filter handling. Previous operating experience with these components in the PASSAT facility

    International Nuclear Information System (INIS)

    In a prototype filter offgas cleaning system for reprocessing plants (PASSAT) built at the Karlsruhe Nuclear Research Center a fullscale filter cell with remotely handled filters for aerosol and iodine removal and the corresponding remote handling systems for exchange, bagging out, packaging and disposal of spent filter elements has been installed and run in trial operation since July 1978. The filters and the replacement techniques have been tested for the past two years or so and so far have always worked satisfactory over the test period involving some 150 replacement events. Neither wear nor corrosion phenomena were found in the filter housings and the replacement systems. The seals and clamping devices were selected so that during operation the prescribed leak rates of -3 Torr l/s were always maintained on the filter lid, the seat of the filter element and the cell lock. The total clamping loads for the filter element and the filter lid amount to approx. 20 kN. The force necessary to separate the filter element from the filter housing is approx. 3.5 kN. No ruptures of seals or gaskets were to be detected. The design of the filters and of the handling systems has been found satisfactorily in the cold test operation so far and can be recommended for use in nuclear facilities. In all experiments conducted until now PASSAT has worked without any failure. All operating data required in the specifications were met in the test period. The maximum pressure loss in the system with loaded filter elements amounts to some 3000 mm of water. After operation with iodine and NO/sub x/, plant components exposed to 100% relative humidity and condensate showed corrosion

  4. Enhancement of the use of digital mock-ups in the verification and validation process for ITER remote handling systems

    Energy Technology Data Exchange (ETDEWEB)

    Sibois, R., E-mail: romain.sibois@vtt.fi [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland); Salminen, K.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland); Mattila, J. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T. [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland)

    2013-10-15

    Highlights: • Verification and validation process for ITER remote handling system. • Verification and validation framework for complex engineering systems. • Verification and validation roadmap for digital modelling phase. • Importance of the product life-cycle management in the verification and validation framework. -- Abstract: The paper is part of the EFDA's programme of European Goal Oriented Training programme on remote handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. This paper is written based on the results of a project “verification and validation (V and V) of ITER RH system using digital mock-ups (DMUs)”. The purpose of this project is to study efficient approach of using DMU for the V and V of the ITER RH system design utilizing a system engineering (SE) framework. This paper reviews the definitions of DMU and virtual prototype and overviews the current trends of using virtual prototyping in the industry during the early design phase. Based on the survey of best industrial practices, this paper proposes ways to improve the V and V process for ITER RH system utilizing DMUs.

  5. Enhancement of the use of digital mock-ups in the verification and validation process for ITER remote handling systems

    International Nuclear Information System (INIS)

    Highlights: • Verification and validation process for ITER remote handling system. • Verification and validation framework for complex engineering systems. • Verification and validation roadmap for digital modelling phase. • Importance of the product life-cycle management in the verification and validation framework. -- Abstract: The paper is part of the EFDA's programme of European Goal Oriented Training programme on remote handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. This paper is written based on the results of a project “verification and validation (V and V) of ITER RH system using digital mock-ups (DMUs)”. The purpose of this project is to study efficient approach of using DMU for the V and V of the ITER RH system design utilizing a system engineering (SE) framework. This paper reviews the definitions of DMU and virtual prototype and overviews the current trends of using virtual prototyping in the industry during the early design phase. Based on the survey of best industrial practices, this paper proposes ways to improve the V and V process for ITER RH system utilizing DMUs

  6. Remote handling equipment for removal of waste from single-shell tanks at the Hanford Site

    International Nuclear Information System (INIS)

    Mechanical retrieval equipment concepts are being developed for remote 'mining' of the radioactive waste in underground storage tanks at the Hanford Site in Washington State. This paper presents a description of the tanks, the waste, the key design considerations, and some of the more promising concepts for mechanical waste retrieval. 6 refs., 4 figs

  7. Remote handling equipment to cut metallic tubes and plates in dangerous atmosphere

    International Nuclear Information System (INIS)

    The aim of the present invention is a remote control pyrotechnic device to cut plates and tubes, more particularly for mechanical piece dismantling in contaminated atmosphere. The present device is compact, easy to use inexpensive and can be disposed of with metallic wastes after its utilization. The device is described in detail, its operation way is also presented

  8. Remote handling prospects. Computer aided remote handling

    International Nuclear Information System (INIS)

    Mechanical manipulators, electrical control manipulators and computer aided manipulators were successively developed. The aim of computer aided manipulators is the realization of complex or tricky job in adverse environment but man is required for non routine work or for situation in evolution. French effort is developed in the frame of the project automation and advanced robotics and new problems have to be solved particularly at the interface man/machine

  9. A practical experience of using special remote handling tools on JET

    International Nuclear Information System (INIS)

    Over 50 cutting and 200 UHV welding operations were made during the installation of new water cooled belt limiters and ICRF Antennae into the JET Vacuum Vessel. This work was performed by the hands-on use of 45 special tools which have been specifically designed for use with the Mascot servomanipulator in preparation for the JET D-T phase when all maintenance will be performed remotely. This paper reports on the techniques used and the performance of the tools. (author)

  10. Electromechanical three-axis development for remote handling in the Hot Experimental Facility

    International Nuclear Information System (INIS)

    A three-axis closed-loop position control system has been designed and installed on an overhead bridge, carriage, tube hoist for automotive positioning of manipulation at a remotely maintained work site. The system provides accurate (within 3 min) and repeatable three-axis positioning of the manipulator. The position control system has been interfaced to a supervisory minicomputer system that provides teach-playback capability of manipulator positioning and color graphic display of the three-axis system position

  11. Remote handling requirements and considerations for D-T fusion reactors

    International Nuclear Information System (INIS)

    This paper an overview of the maintenance considerations of next-generation fusion reactors. It draws upon the work done at the Fusion Engineering Design Center over the past several years in the conceptual development of tokamaks and tandem mirrors. It specifically addresses the maintenance philosophy adopted for these devices, the configuration development using a modular design approach, scheduled and unscheduled maintenance operations, assembly and disassembly scenarios for component replacements, maintenance equipment requirements, and the operating availability of these devices. In addition, recent work on the development of a totally remote tokamak configuration is presented

  12. Cultural Resource Protection Plan for the Remote-Handled Low-Level Waste Disposal Facility at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Pace, Brenda Ringe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gilbert, Hollie Kae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This plan addresses cultural resource protection procedures to be implemented during construction of the Remote Handled Low Level Waste project at the Idaho National Laboratory. The plan proposes pre-construction review of proposed ground disturbing activities to confirm avoidance of cultural resources. Depending on the final project footprint, cultural resource protection strategies might also include additional survey, protective fencing, cultural resource mapping and relocation of surface artifacts, collection of surface artifacts for permanent curation, confirmation of undisturbed historic canal segments outside the area of potential effects for construction, and/or archaeological test excavations to assess potential subsurface cultural deposits at known cultural resource locations. Additionally, all initial ground disturbing activities will be monitored for subsurface cultural resource finds, cultural resource sensitivity training will be conducted for all construction field personnel, and a stop work procedure will be implemented to guide assessment and protection of any unanticipated discoveries after initial monitoring of ground disturbance.

  13. Use of curium spontaneous fission neutrons for safeguardability of remotely-handled nuclear facilities: Fuel fabrication in pyroprocessing

    International Nuclear Information System (INIS)

    Highlights: • The neutron flux due to the metal fuel alloy is modeled. • This is part of a series of studies into major system components. • The strength of the neutron flux will affect hot cell shielding and design. • A residual mass in the injection casting system was determined. • This residual mass could be used in materials accounting. -- Abstract: Advanced nuclear reactor systems (NESs) will utilize remotely-handled facilities in which batch-type processing will occur in hot cells. There are no current formalized criteria for International Atomic Energy Agency (IAEA) safeguards for these systems. This creates new challenges to develop methodologies for demonstrating the safeguardability of these facilities. A High Reliability Safeguards (HRS) approach therefore has been proposed to enhance intrinsic proliferation resistance by establishing an envelope of adaptable functional components as part of a facility design strategy. Additionally, system assessment can be modeled concurrently with safety and physical security by a risk-informed approach. The HRS approach is currently applied to a commercial pyroprocessing facility as an example system. A scoping study is presented as the first in a series of quantitative modeling efforts to extend the HRS approach. These efforts currently focus on investigating the magnitude of neutron fluxes due to spontaneous fission of curium for commercial batch sizes and held up materials for important processes in the system. Here, the fuel fabrication process is studied. The intent of these initial studies is to learn how the intrinsic properties of materials in the pyroprocessing system will affect facility design and safeguards. The model presented in this paper is intended to be adaptable to more practical and complex scenarios in order to evaluate the safeguardability of remotely-handled nuclear facilities

  14. Status of microwave process development for RH-TRU [remote-handled transuranic] wastes at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    The Oak Ridge National Laboratory (ORNL) Waste Handling and Packaging Plant is developing a microwave process to reduce and solidify remote-handled transuranic (RH-TRU) liquids and sludges presently stored in large tanks at ORNL. Testing has recently begun on an in-drum microwave process using nonradioactive RH-TRU surrogates. The microwave process development effort has focused on an in-drum process to dry the RH-TRU liquids and sludges in the final storage container and then melt the salt residues to form a solid monolith. A 1/3-scale proprietary microwave applicator was designed, fabricated, and tested to demonstrate the essential features of the microwave design and to provide input into the design of the full-scale applicator. The microwave fields are uniform in one dimension to reduce the formation of hot spots on the microwaved wasteform. The final wasteform meets the waste acceptance criteria for the Waste Isolation Pilot Plant, a federal repository for defense transuranic wastes near Carlsbad, New Mexico. 7 refs., 1 fig., 1 tab

  15. Irradiation and testing of off-the-shelf seal materials for water hydraulic applications in ITER remote handling equipment

    International Nuclear Information System (INIS)

    Remote handling (RH) is one of the most challenging aspects of the ITER project, and the European home team is building a major prototype of the divertor region (the Divertor Test Platform 2) to confirm practically the RH concepts proposed in this area. To handle the 9 Tonne divertor cassette, water hydraulics has been selected because it offers high forces and precise control in a compact envelope, with minimal long-term contamination should a leak develop. Water hydraulic components use mainly stainless steel - unaffected by gamma radiation - but the integral seals and O-rings are known to be sensitive. For radiation testing of these components, a modular approach was adopted, enabling up to 11 seal carriers assemblies to be irradiated simultaneously in the limited space available, with individual carriers being removed at varying total doses up to 10 MGy. Each carrier was then installed in a real hydraulic rig for testing, revealing not only at what total dose the components became unusable, but also how they fail, enabling condition monitoring to assess the state of the seals long before their failure might render the RH equipment irrecoverable

  16. Radiological risk assessment for the remote-handled transuranic waste storage options at Argonne National Laboratory - East

    International Nuclear Information System (INIS)

    Interim storage of the remote-handled transuranic (RH/TRU) waste is needed at Argonne National Laboratory-East (ANL-E). Two on-site facilities, the northwest (NW) vaults in the 317 Area and the converted spent nuclear fuel pool in Building 331, were identified as potential storage locations through previous studies. To assist the decision making process of selecting a storage location, radiological risk assessments were conducted to analyze potential radiation exposures that would be associated with storage of the RH/TRU waste in these two facilities. Three drum storage scenarios (one for the 317 Area and two for Building 331) considering different drum handling procedures and stacking patterns were developed. Time-motion information on worker activities that would occur in the procedures was collected and recorded in spreadsheets. Using the time-motion information, potential external doses were estimated for the involved workers for each step in the procedures. The sum of the potential external doses over all the activity steps gave the total collective dose for each scenario. The results show that during the storage phase, storing waste drums in half-liners in Building 331 would result in the lowest collective radiation exposure; however, it would also require the most human resources. When retrieving waste drums for off-site shipment was considered, storing waste drums in the 317 Area would be the most favorable option because it would require the least amount of human resources and would also result in the lowest collective radiation exposure

  17. A remote handling rate-position controller for telemanipulating in a large workspace

    Energy Technology Data Exchange (ETDEWEB)

    Barrio, Jorge, E-mail: jordi.barrio@upm.es; Ferre, Manuel, E-mail: m.ferre@upm.es; Suárez-Ruiz, Francisco, E-mail: fa.suarez@upm.es; Aracil, Rafael, E-mail: rafael.aracil@upm.es

    2014-01-15

    This paper presents a new haptic rate-position controller, which allows manipulating a slave robot in a large workspace using a small haptic device. This control algorithm is very effective when the master device is much smaller than the slave device. Haptic information is displayed to the user so as to be informed when a change in the operation mode occurs. This controller allows performing tasks in a large remote workspace by using a haptic device with a reduced workspace such as Phantom. Experimental results have been carried out using a slave robot from Kraft Telerobotics and a commercial haptic interface as a master device. A curvature path following task has been simulated using the proposed controller which was compared with the force-position control algorithm. Results obtained show that higher accuracy is obtained when the proposed method is used, spending a similar amount of time to perform the task.

  18. A remote handling rate-position controller for telemanipulating in a large workspace

    International Nuclear Information System (INIS)

    This paper presents a new haptic rate-position controller, which allows manipulating a slave robot in a large workspace using a small haptic device. This control algorithm is very effective when the master device is much smaller than the slave device. Haptic information is displayed to the user so as to be informed when a change in the operation mode occurs. This controller allows performing tasks in a large remote workspace by using a haptic device with a reduced workspace such as Phantom. Experimental results have been carried out using a slave robot from Kraft Telerobotics and a commercial haptic interface as a master device. A curvature path following task has been simulated using the proposed controller which was compared with the force-position control algorithm. Results obtained show that higher accuracy is obtained when the proposed method is used, spending a similar amount of time to perform the task

  19. Remote handling dynamical modelling: assessment on new approach to enhance positioning accuracy with heavy load manipulation

    International Nuclear Information System (INIS)

    In vessel maintenance work in Fusion Tokamak will be carried out with help several sets of robotic devices. Heavy loads handling in constrained space is identified by all players of the RH community as a key-issue in the latest Fusion Tokamak facilities. To deal with high-level dexterity tasks, high payload to mass ratio and limited operating space, RH equipment designers can only propose systems whose mechanical flexibility is no longer negligible and need to be taken into account in the control scheme. Traditional approaches where control system only includes a linear model of deformation of the structure leads to poor positioning accuracy. Uncontrolled or under evaluated errors could be damaging for in-vessel components during maintenance operations in the Tokamak facility. To address the control of complex flexible systems, we will investigate the use of specific mechanical software that combines both finite element and kinematical joints analyses, with a strong-coupled formulation, to perform system dynamics simulations. This procedure will be applied on a single axis mock up robotic joint with highly flexible structure. A comparison of experimental results with the traditional linear approach and the specified software model will be carried out. Benefits introduced by this new approach will finally be assessed in view of RH design or specification in the field of RH in Fusion Tokamak scale such as ITER. (orig.)

  20. Automation, robotics and remote handling technology in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Automation and Robotics technology are making significant contributions in almost all fields of engineering and technology and their presence is felt in all spheres of human life. The importance of automation and robotics has increased rapidly in the recent years to cater to the global competitive pressures by the manufacturing industry by utilizing the increased productivity and improved quality this technology offers. Improvement of productivity, quality, profitability and, indeed, survival are the major motivating factors in the implementation of automation and robotics technology in the manufacturing sector. Robots are used extensively in the automotive industry, primarily for welding, painting and material handling applications. The electronics, aerospace, metalworking and consumer goods industries are also major potential robot users. The common uses of robots in industries mostly include the four Ps - Picking, Placing, Packaging and Painting - apart from other industrial routines like assembly and welding. As is the case with industrial tools and machineries, a properly designed robot (for the appropriate task) has almost unlimited endurance with the added benefit of precisions unmatched by human workers. With robot technology as a key element, integrated factory automation systems touch on nearly all types of manufacturing. The productivity and competitiveness in these industries will depend in large part on flexible automation through robotics

  1. Development of an Anthropomorphous, Remote-Controlled Machine for Handling Radiation Accidents

    International Nuclear Information System (INIS)

    In handling radiation accidents personnel should be kept outside dangerous radiation fields wherever possible. It is obvious that human beings should generally be replaced by machines, which should possess the universal attributes of man. but be insensitive to high radiation activities. The master-slave system appears especially suitable. In this sytem the master is a human being who stays away from the radiation field in a safe room equipped with the necessary information and telecommand devices, and controls and watches the work of the slave from there. The master represents the brain of the machine. The slave is the actual machine and to a certain extent the executive motor part of the human being. It is equipped with limb-like devices, some sense organs such as stereo-television eyes, stereo-microphones, and, by means of power force meters, possesses a certain sense of touch. Further development towards a biped walking machine is outlined. The machine must be given human dimensions as existing atomic plants are made for human operation. (author)

  2. Remote handling dynamical modelling: assessment on new approach to enhance positioning accuracy with heavy load manipulation

    Energy Technology Data Exchange (ETDEWEB)

    Gagarina-Sasia, T.; David, O.; Dubus, G.; Perrot, Y.; Riwain, A. [CEA, LIST, 92 - Fontenay aux Roses (France). Service de Robotique Interactive; Gabellini, E.; Pretot, P. [SAMTECH, 91 - Massy (France)

    2007-07-01

    In vessel maintenance work in Fusion Tokamak will be carried out with help several sets of robotic devices. Heavy loads handling in constrained space is identified by all players of the RH community as a key-issue in the latest Fusion Tokamak facilities. To deal with high-level dexterity tasks, high payload to mass ratio and limited operating space, RH equipment designers can only propose systems whose mechanical flexibility is no longer negligible and need to be taken into account in the control scheme. Traditional approaches where control system only includes a linear model of deformation of the structure leads to poor positioning accuracy. Uncontrolled or under evaluated errors could be damaging for in-vessel components during maintenance operations in the Tokamak facility. To address the control of complex flexible systems, we will investigate the use of specific mechanical software that combines both finite element and kinematical joints analyses, with a strong-coupled formulation, to perform system dynamics simulations. This procedure will be applied on a single axis mock up robotic joint with highly flexible structure. A comparison of experimental results with the traditional linear approach and the specified software model will be carried out. Benefits introduced by this new approach will finally be assessed in view of RH design or specification in the field of RH in Fusion Tokamak scale such as ITER. (orig.)

  3. Tritium contamination studies involving test materials and jet remote handling tools

    International Nuclear Information System (INIS)

    To determine the potential contamination of remote cutting and welding tools to be used in the JET torus after the introduction of tritium, experiments were performed using these tools on INCONEL pipe specimens which had been exposed to elemental tritium (HT) at a concentration of 4.6 x 1010 Bq/m3. A maximum tritium release of ∼15,600 Bq was measured during welding, resulting in the tool's surface contamination of 0.5 Bq/cm2. A second series of tests was performed in order to determine the degree of surface contamination of various materials when exposed to HTO as a function of the exposure time and the relative efficacy of different decontamination techniques. Stainless steel, aluminium alloy and PVC rigid were exposed to HTO (liquid) at a concentration 4.4 x 1010 Bq/1 for 1, 24, 120 hours and decontaminated. The decontamination techniques used included; leaching in water, baking at 100 degree C, hot air stream, weathering. The maximum levels of tritium surface contamination measured during the test were ∼12 Bq/cm2 for stainless steel, ∼ Bq/cm2 for aluminium alloy and ∼1,700 Bq/cm2 for PVC. A decontamination factor of about 80% as measured by smears was achieved using hot air stream at 125 degree C on stainless steel and aluminium alloy and baking PVC at 100 degree C. 6 figs., 2 tabs

  4. Breeding blanket for DEMO

    International Nuclear Information System (INIS)

    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 being investigated within the framework of the European Test-Blanket Development Programme. (orig.)

  5. IFMIF – Layout and arrangement of cells according to requirements of technical logistics, reliability and remote handling

    International Nuclear Information System (INIS)

    Highlights: ► In a first approach, layout and arrangement of the cells followed a predetermined plant layout. ► Disadvantages in technical logistics, reliability and remote handling have been detected. ► Deliberation with project teams opened space for improvements. ► Layout and arrangement of cells have been improved by simplification of design. ► Speed and reliability have been increased significantly. - Abstract: The International Fusion Material Irradiation Facility (IFMIF) is designed to study and qualify structural and functional materials which shall be used in future fusion nuclear power plants. During the current engineering validation and engineering design activities (EVEDA) phase the development of e.g. an optimized layout and arrangement of the cells (Access Cell, Test Cell, and Test Module Handling Cells) is of major interest. After defining different functions for the individual cells like e.g. large scale/fine scale disassembling of test modules a first layout has been developed. This design followed requirements like having a minimum of carrier changes to avoid sources of failures. On the other hand it has had to be a compact arrangement of cells due to restrictions from plant layout. A row of changes of transfer direction, and different crane systems were the consequence. Constructive discussion with project team results in the statement, that for reasons of being reliable and fast, layout and arrangement of cells goes first, plant layout then will follow. The chance for big improvements was taken and the result was a simplified design with strong reduced number of functional elements, and increased reliability and speed.

  6. Remote handling techniques in decommissioning - A report of the NEA Co-operative Programme on Decommissioning (CPD) project

    International Nuclear Information System (INIS)

    The NEA Co-operative Programme for the Exchange of Scientific and Technical Information Concerning Nuclear Installation Decommissioning Projects (CPD) is a joint undertaking of a limited number of organisations actively executing on planning the decommissioning of nuclear facilities. The objective of the CPD is to acquire information from operational experience in decommissioning nuclear installations that is useful for future projects. Although part of the information exchanged within CPD is confidential in nature and is restricted to programme participants, experience of general interest gained under the programme's auspices is released for broader use. Such information is brought to the attention of all NEA members through regular reports to the NEA Radioactive Waste Management Committee (RWMC), as well as through published studies. This report describes generic results obtained by a CPD Task Group analysing the needs for remote technologies. The existing technologies able to meet these needs, the lessons learned and showing where improvements or further developments should be made in this domain. During the D and D process, the handling of highly radioactive materials, the deployment of tools and sensors and the dismantling of components built from many different materials can be a long, labor-intensive process that has the potential for high exposure rates, heat stress and injury to personnel. Mobile robotics systems provide solutions to these hazards. Such remote handling systems are required to perform tasks within budget and on schedule while justifying the expense by a saving in cumulative doses received by project personnel. To reach this goal, the following are additional factors that need to be evaluated when preparing a project: - System and peripherals must be operator-friendly. Ideally, the system must be designed to allow personnel currently available for the D and D project to become trained as operators within a reasonable time frame. - The

  7. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    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

  8. An electro-hydraulic servo control system research for CFETR blanket RH

    International Nuclear Information System (INIS)

    Highlights: • We discussed the conceptual design of CFETR blanket RH maintenance system. • The mathematical model of electro-hydraulic servo system was calculated. • A fuzzy adaptive PD controller was designed based on control theory and experience. • The co-simulation models of the system were established with AMESim/Simulink. • The fuzzy adaptive PD algorithm was designed as the core strategy of the system. - Abstract: Based on the technical design requirements of China Fusion Engineering Test Reactor (CFETR) blanket remote handling (RH) maintenance, this paper focus on the control method of achieving high synchronization accuracy of electro-hydraulic servo system. Based on fuzzy control theory and practical experience, a fuzzy adaptive proportional-derivative (PD) controller was designed. Then a more precise co-simulation model was established with AMESim/Simulink. Through the analysis of simulation results, a fuzzy adaptive PD control algorithm was designed as the core strategy of electro-hydraulic servo control system

  9. Preconceptual engineering design for the APT 3He Target/Blanket concept

    International Nuclear Information System (INIS)

    A preconceptual engineering design has been developed for the 3He Target/Blanket (T/B) System for the Accelerator Production of Tritium Project. This concept uses an array of pressure tubes containing tungsten rods for the neutron spallation source and 3He gas contained in a metal tank and blanket tubes as the tritium production material. The engineering design is based on a physics model optimized for efficient tritium production. Principle engineering consideration were: provisions for cooling all materials including the 3He gas; containment of the gas and radionuclides; remote handling; material compatibility; minimization of 3He, D2O, and activated waste; modularity; and manufacturability. The design provides a basis for estimating the cost to implement the system

  10. An electro-hydraulic servo control system research for CFETR blanket RH

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Changqi [Hefei University of Technology, Hefei 230009, Anhui (China); Tang, Hongjun, E-mail: taurustang@126.com [Hefei University of Technology, Hefei 230009, Anhui (China); Qi, Songsong [Hefei University of Technology, Hefei 230009, Anhui (China); Cheng, Yong; Feng, Hansheng; Peng, Xuebing; Song, Yuntao [Institute of Plasma Physics Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2014-11-15

    Highlights: • We discussed the conceptual design of CFETR blanket RH maintenance system. • The mathematical model of electro-hydraulic servo system was calculated. • A fuzzy adaptive PD controller was designed based on control theory and experience. • The co-simulation models of the system were established with AMESim/Simulink. • The fuzzy adaptive PD algorithm was designed as the core strategy of the system. - Abstract: Based on the technical design requirements of China Fusion Engineering Test Reactor (CFETR) blanket remote handling (RH) maintenance, this paper focus on the control method of achieving high synchronization accuracy of electro-hydraulic servo system. Based on fuzzy control theory and practical experience, a fuzzy adaptive proportional-derivative (PD) controller was designed. Then a more precise co-simulation model was established with AMESim/Simulink. Through the analysis of simulation results, a fuzzy adaptive PD control algorithm was designed as the core strategy of electro-hydraulic servo control system.

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

    International Nuclear Information System (INIS)

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

  12. Assessment of Geochemical Environment for the Proposed INL Remote-Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    D. Craig Cooper

    2011-11-01

    Conservative sorption parameters have been estimated for the proposed Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility. This analysis considers the influence of soils, concrete, and steel components on water chemistry and the influence of water chemistry on the relative partitioning of radionuclides over the life of the facility. A set of estimated conservative distribution coefficients for the primary media encountered by transported radionuclides has been recommended. These media include the vault system, concrete-sand-gravel mix, alluvium, and sedimentary interbeds. This analysis was prepared to support the performance assessment required by U.S. Department of Energy Order 435.1, 'Radioactive Waste Management.' The estimated distribution coefficients are provided to support release and transport calculations of radionuclides from the waste form through the vadose zone. A range of sorption parameters are provided for each key transport media, with recommended values being conservative. The range of uncertainty has been bounded through an assessment of most-likely-minimum and most-likely-maximum distribution coefficient values. The range allows for adequate assessment of mean facility performance while providing the basis for uncertainty analysis.

  13. Enhancement of the remote handling strategy for the refurbishment of the backplate bayonet concept of IFMIF target system

    Energy Technology Data Exchange (ETDEWEB)

    Micciche, G., E-mail: gioacchino.micciche@enea.it [CR ENEA Brasimone, I-40035 Camugnano, BO (Italy); Lorenzelli, L.; Bernardi, D. [CR ENEA Brasimone, I-40035 Camugnano, BO (Italy); Queral, V. [EURATOM-CIEMAT, Avda. Computense 22, 28040 Madrid (Spain)

    2011-10-15

    One of the most technically challenging activities of the IFMIF facility is the maintenance and the refurbishment of its components, and among these the target system appears to be critical since it is located in the most severe region of neutron irradiation (60 dpa/fpy). Two different target assembly systems have been developed: the first is known as integral target while the second one is based on the so called replaceable backplate bayonet concept. The present remote handling (RH) procedures developed for the refurbishment of the removable backplate foresee the removal of all the components from the upper part of the test cell. This operation has a strong impact on the intervention time for the backplate refurbishment which has to be repeated at least every year. Consequently the need to update the RH strategy for the refurbishment of this component becomes a precondition in order to fulfill the stringent requirement to enhance the duty cycle of IFMIF plant. In this paper two potential approaches are presented: the first relies on the possibility to perform all the refurbishment operations in situ in the test cell cavern, whilst the second one foresees to perform these operations off-line in a hot cell. Advantages and disadvantages of these approaches together with the RH requirements for the refurbishment operations of the backplate bayonet concepts are also reported.

  14. Materials for breeding blankets

    International Nuclear Information System (INIS)

    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

  15. Structural design and preliminary analysis of liquid lead–lithium blanket for China Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    China Fusion Engineering Test Reactor (CFETR) has been proposed as an option in China to bridge the gaps between ITER and fusion power plant. Since one major goal of CFETR is to demonstrate long pulse or steady-state operation with duty cycle time ≥0.3–0.5, easier maintenance of the in-vessel components is emphasized in the design process. In this contribution, a kind of liquid lead–lithium tritium breeder blanket concept focus on the remote maintenance has been designed for CFETR. To make the pipes and mechanical connections at the rear of the blanket accessible from vacuum vessel, two kinds of guide tubes were adopted to provide passageways for remote handling tools. In order to evaluate the effects of the guide tube installation on the structural performance of the blanket, as a preliminary stage, thermal-hydraulic analysis of first wall was carried out based on the heat load obtained from 3D modeled neutronics calculations. In addition, thermal stress analysis of the first wall under normal condition was performed to evaluate the thermomechanical behavior. The preliminary analysis results validated the performance of current blanket design

  16. Structural design and preliminary analysis of liquid lead–lithium blanket for China Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ni, Muyi; Lian, Chao; Zhang, Shichao; Nie, Baojie; Jiang, Jieqiong, E-mail: jieqiong.jiang@fds.org.cn

    2015-05-15

    China Fusion Engineering Test Reactor (CFETR) has been proposed as an option in China to bridge the gaps between ITER and fusion power plant. Since one major goal of CFETR is to demonstrate long pulse or steady-state operation with duty cycle time ≥0.3–0.5, easier maintenance of the in-vessel components is emphasized in the design process. In this contribution, a kind of liquid lead–lithium tritium breeder blanket concept focus on the remote maintenance has been designed for CFETR. To make the pipes and mechanical connections at the rear of the blanket accessible from vacuum vessel, two kinds of guide tubes were adopted to provide passageways for remote handling tools. In order to evaluate the effects of the guide tube installation on the structural performance of the blanket, as a preliminary stage, thermal-hydraulic analysis of first wall was carried out based on the heat load obtained from 3D modeled neutronics calculations. In addition, thermal stress analysis of the first wall under normal condition was performed to evaluate the thermomechanical behavior. The preliminary analysis results validated the performance of current blanket design.

  17. Irradiation and Testing of off-the-shelf Seal Materials for Water Hydraulic Applications in ITER Remote Handling Equipment

    International Nuclear Information System (INIS)

    Remote Handling (RH) is one of the most challenging aspects of the ITER project, and the European home team is presently designing and constructing a major prototype of the divertor region (the Divertor Test Platform 2) to confirm practically the concepts proposed for the RH systems in this area. The divertor handling equipment must lift and transport the 9 Tonne divertor cassette around the vessel and down the narrow duct from the vessel to a sealed cask which will take the cassette to the hot cell. Due to limited access, the cassette must be grappled in a cantilevered manner, and water hydraulics has been selected because the required high forces and precise control are available in a compact envelope, with minimal long-term contamination of the vessel or duct should a leak develop. Although the main material used in water hydraulic components is stainless steel (unaffected by secondary gamma radiation), the seals and O-rings inside these components are more sensitive. With promising materials identified in earlier tests, a number of identical seal-carrier assemblies were manufactured upon which standard seals and O-rings were fitted and subjected to increasing doses of irradiation representative of those predicted in the radial duct region. Since the seal carriers are annular, space is also available for material samples for complementary mechanical testing. After irradiation, the carriers can then be assembled onto a specially modified piston and cylinder where various key parameters, like friction and leakage can be derived and compared with the same tests carried out on the pre-irradiated assemblies. This modular approach minimises the space required for irradiation (a chamber 60 mm diameter x 100 mm long can hold 10 seal assemblies), enables the effects of progressive radiation to be studied, and can allow several materials to be tested under realistic hydraulic conditions in a short time. Initial tests used UHMW-PE material for the seals, and NBR for the O

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

    International Nuclear Information System (INIS)

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 106 R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

  20. Pre-conceptual Design Assessment of DEMO Remote Maintenance

    CERN Document Server

    Loving, A; Sykes, N; Iglesias, D; Coleman, M; Thomas, J; Harman, J; Fischer, U; Sanz, J; Siuko, M; Mittwollen, M; others,

    2013-01-01

    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 2012/13, 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 ...

  1. Technology Development And Deployment Of Systems For The Retrieval And Processing Of Remote-Handled Sludge From Hanford K-West Fuel Storage Basin

    International Nuclear Information System (INIS)

    In 2011, significant progress was made in developing and deploying technologies to remove, transport, and interim store remote-handled sludge from the 105-K West Fuel Storage Basin on the Hanford Site in south-central Washington State. The sludge in the 105-K West Basin is an accumulation of degraded spent nuclear fuel and other debris that collected during long-term underwater storage of the spent fuel. In 2010, an innovative, remotely operated retrieval system was used to successfully retrieve over 99.7% of the radioactive sludge from 10 submerged temporary storage containers in the K West Basin. In 2011, a full-scale prototype facility was completed for use in technology development, design qualification testing, and operator training on systems used to retrieve, transport, and store highly radioactive K Basin sludge. In this facility, three separate systems for characterizing, retrieving, pretreating, and processing remote-handled sludge were developed. Two of these systems were successfully deployed in 2011. One of these systems was used to pretreat knockout pot sludge as part of the 105-K West Basin cleanup. Knockout pot sludge contains pieces of degraded uranium fuel ranging in size from 600 μm to 6350 μm mixed with pieces of inert material, such as aluminum wire and graphite, in the same size range. The 2011 pretreatment campaign successfully removed most of the inert material from the sludge stream and significantly reduced the remaining volume of knockout pot product material. Removing the inert material significantly minimized the waste stream and reduced costs by reducing the number of transportation and storage containers. Removing the inert material also improved worker safety by reducing the number of remote-handled shipments. Also in 2011, technology development and final design were completed on the system to remove knockout pot material from the basin and transport the material to an onsite facility for interim storage. This system is scheduled

  2. Assessment of Potential Flood Events and Impacts at INL's Proposed Remote-Handled Low-Level Waste Disposal Facility Sites

    International Nuclear Information System (INIS)

    Rates, depths, erosion potential, increased subsurface transport rates, and annual exceedance probability for potential flooding scenarios have been evaluated for the on-site alternatives of Idaho National Laboratory's proposed remote handled low-level waste disposal facility. The on-site disposal facility is being evaluated in anticipation of the closure of the Radioactive Waste Management Complex at the INL. An assessment of flood impacts are required to meet the Department of Energy's Low-Level Waste requirements (DOE-O 435.1), its natural phenomena hazards assessment criteria (DOE-STD-1023-95), and the Radioactive Waste Management Manual (DOE M 435.1-1) guidance in addition to being required by the National Environmental Policy Act (NEPA) environmental assessment (EA). Potential sources of water evaluated include those arising from (1) local precipitation events, (2) precipitation events occurring off of the INL (off-site precipitation), and (3) increased flows in the Big Lost River in the event of a Mackay Dam failure. On-site precipitation events include potential snow-melt and rainfall. Extreme rainfall events were evaluated for the potential to create local erosion, particularly of the barrier placed over the disposal facility. Off-site precipitation carried onto the INL by the Big Lost River channel was evaluated for overland migration of water away from the river channel. Off-site precipitation sources evaluated were those occurring in the drainage basin above Mackay Reservoir. In the worst-case scenarios, precipitation occurring above Mackay Dam could exceed the dam's capacity, leading to overtopping, and eventually complete dam failure. Mackay Dam could also fail during a seismic event or as a result of mechanical piping. Some of the water released during dam failure, and contributing precipitation, has the potential of being carried onto the INL in the Big Lost River channel. Resulting overland flows from these flood sources were evaluated for their

  3. Limitations on blanket performance

    International Nuclear Information System (INIS)

    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.)

  4. Research on the market availability of mining machines with specific features (radiation protection, remote handling equipment, electric power drive) for the radioactive waste retrieval

    International Nuclear Information System (INIS)

    The report has been prepared in the frame of the required fact assessment process concerning the state of the waste packages in the emplacement cabins of the Schachtanlage Asse. The search on the market availability of mining machines with specific features (radiation protection, remote handling equipment, electric power drive) for the radioactive waste retrieval covered the following issues: gangue exploitation and room-and-pillar work, underground hauling and transport, railless special vehicles for underground transportation, continuous conveyors and components, feeding and removal equipment, control techniques communication and navigation. The report includes a summary of results and open questions.

  5. Gamma-ray spectrometry combined with acceptable knowledge (GSAK). A technique for characterization of certain remote-handled transuranic (RH-TRU) wastes. Part 1. Methodology and techniques

    International Nuclear Information System (INIS)

    Gamma-ray spectrometry combined with acceptable knowledge (GSAK) is a technique for the characterization of certain remote-handled transuranic (RH-TRU) wastes. GSAK uses gamma-ray spectrometry to quantify a portion of the fission product inventory of RH-TRU wastes. These fission product results are then coupled with calculated inventories derived from acceptable process knowledge to characterize the radionuclide content of the assayed wastes. GSAK has been evaluated and tested through several test exercises. GSAK approach is described, while test results are presented in Part II. (author)

  6. Gamma-ray spectrometry combined with acceptable knowledge (GSAK). A technique for characterization of certain remote-handled transuranic (RH-TRU) wastes. Part 2. Testing and results

    International Nuclear Information System (INIS)

    Gamma-ray spectrometry combined with acceptable knowledge (GSAK) is a technique for the characterization of certain remote-handled transuranic (RH-TRU) wastes. GSAK uses gamma-ray spectrometry to quantify a portion of the fission product inventory of RH-TRU wastes. These fission product results are then coupled with calculated inventories derived from acceptable process knowledge to characterize the radionuclide content of the assayed wastes. GSAK has been evaluated and tested through several test exercises. These tests and their results are described; while the former paper in this issue presents the methodology, equipment and techniques. (author)

  7. The use of virtual reality and intelligent database systems for procedure planning, visualisation, and real-time component tracking in remote handling operations

    International Nuclear Information System (INIS)

    The organisation of remote handling (RH) operations in fusion environments is increasingly critical as the number of tasks, components and tooling that RH operations teams must deal with inexorably rises. During the recent JET EP1 RH shutdown the existing virtual reality (VR) and procedural database systems proved essential for visualisation and tracking of operations, particularly due to the increasing complexity of remote tasks. A new task planning system for RH operations is in development, and is expected to be ready for use during the next major shutdown, planned for 2009. The system will make use of information available from the remote operations procedures, the RH equipment human-machine interfaces, the on-line RH equipment control systems and also the virtual reality (VR) system to establish a complete database for the location of plant items and RH equipment as RH operations progress. It is intended that the system be used during both preparation and implementation of shutdowns. In the preparations phase the system can be used to validate procedures and overall logistics by allowing an operator to increment through each operation step and to use the VR system to visualise the location and status of all components, manipulators and RH tools. During task development the RH operations engineers can plan and visualise movement of components and tooling to examine handling concepts and establish storage requirements. In the implementation of operations the daily work schedules information will be integrated with the RH operations procedures tracking records to enable the VR system to provide a visual representation of the status of remote operations in real time. Monitoring of the usage history of items will allow estimates of radiation dosage and contaminant exposure to be made. This paper describes the overall aims, structure and use of the system, discusses its application to JET and also considers potential future developments.

  8. Solid breeder blanket concepts

    International Nuclear Information System (INIS)

    An investigation is made of a mechanical concept for the blanket with solid breeders in view of the possible adaptation to power reactor. A special arrangement of the multiplier and breeder materials is developed to permit a further neutronic optimisation

  9. Overview of remote-maintenance scenarios for the ITER machine

    International Nuclear Information System (INIS)

    Maintenance of the International Thermonuclear Experimental Reactor (ITER) will have to be carried out remotely. A preliminary study has been made of remote-handling scenarios of the main components, including blanket, divertor and coils. Frequent scheduled maintenance operations will be carried out without breaking the cryostat vacuum and by working from (shielded) containers connected to maintenance ports external to the cryostat. Exchange of the blanket is foreseen after the initial basic performance phase. This involves application of special welding and cutting techniques that will have to be developed and remotized, as well as handling of modules weighing up to 60 tonnes via complex trajectories inside the vessel and through narrow ports, whilst balancing forces will have to be applied to counteract the out-of-centre-of-gravity lifting. Maintenance scenarios are designed with radiation exposure and contamination control as an overriding requirement. This may require the use of shielded, very heavy containment casks. Following the detailed study of remote handling feasibility, equipment design will proceed up to the end of Engineering Design Activities and beyond. Development will be required for welding, cutting and inspection equipment, and for radiation-hard components, and tests will have to be undertaken to verify particularly difficult operations. (orig.)

  10. Conceptual design of solid breeder blanket system cooled by supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio; Akiba, Masato [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment] [and others

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li{sub 2}TiO{sub 3} or Li{sub 2}O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for

  11. Conceptual design of solid breeder blanket system cooled by supercritical water

    International Nuclear Information System (INIS)

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li2TiO3 or Li2O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for the energy

  12. First analysis of remote handling maintenance procedure in the hot cell for the ITER ICH and CD antenna – RVTL replacement

    International Nuclear Information System (INIS)

    This paper deals with a first analysis of the Remote Handling (RH) maintenance procedure for the replacement of Removable Vacuum Transmission Lines (RVTL) of the ICRH antenna Port Plug (PP). In the framework of the grant F4E-2009-GRT-026, CEA IRFM studied the maintenance in parallel with the design of the antenna provided by CCFE. The RVTL are 8 components of the ICRH antenna which form the interface between the matching system and the four port junction integrating the straps. A folded stub is attached to the principal line to ensure water cooling of the interspace. At the front and the rear of the RVTL are installed double RF windows that provide the first tritium barrier. In case of failure of the first window, all the RVTL have to be replaced. Due to the contamination and activation, the replacement must take place in the hot cell. The complete maintenance sequence is studied. It starts when the PP is in place in the Tokamak equatorial port, it continues with: the preparation in the port cell, transfer to the HC, cleaning, RVTL replacement, returns to the port cell. It finishes with reconnection to the port. The ITER requirements [1] and the hot cell constraints [2] are used to extract specifications for the RH tooling (for handling, cutting, welding, etc.). Each step is studied and suitable tools identified. For specific steps, mechanical concepts for dedicated tools are proposed. Furthermore, the critical steps identified are simulated to check the feasibility

  13. Diffusive heat blanketing envelopes of neutron stars

    CERN Document Server

    Beznogov, M V; Yakovlev, D G

    2016-01-01

    We construct new models of outer heat blanketing envelopes of neutron stars composed of binary ion mixtures (H - He, He - C, C - Fe) in and out of diffusive equilibrium. To this aim, we generalize our previous work on diffusion of ions in isothermal gaseous or Coulomb liquid plasmas to handle non-isothermal systems. We calculate the relations between the effective surface temperature Ts and the temperature Tb at the bottom of heat blanketing envelopes (at a density rhob= 1e8 -- 1e10 g/cc) for diffusively equilibrated and non-equilibrated distributions of ion species at different masses DeltaM of lighter ions in the envelope. Our principal result is that the Ts - Tb relations are fairly insensitive to detailed distribution of ion fractions over the envelope (diffusively equilibrated or not) and depend almost solely on DeltaM. The obtained relations are approximated by analytic expressions which are convenient for modeling the evolution of neutron stars.

  14. Neutron shielding analysis for remote handled transuranic waste containers in facility casks at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Neutron shielding characteristics of the Waste Isolation Pilot Plant facility cask have been quantified for a variety of combinations of neutron sources and waste matrices which would potentially be handled in waste containers. The neutron attenuation and neutron environment of the waste container and the facility cask have been analyzed to ensure that the design requirement of neutron dose rate will be met under the combinations of the source and waste matrix conditions. The analyses considered the ranges of neutron source spectrum and waste matrices which combine to produce the minimum neutron shielding worth of the facility cask. One-dimensional analyses were performed with discrete ordinate transport theory methods using multigroup neutron cross section data. The results discussed in this report demonstrate the effect of source spectrum and waste container matrix on predicted neutron dose rates adjacent to the unshielded waste container and the surface of the facility cask. An evaluation of the uncertainties in predicted neutron dose rates is provided which results in an assessment of the maximum measured neutron dose rate external to the facility cask. A description of the analytical models developed, the analysis methodology, the neutron source spectra, and the detailed results are described in this report. 10 refs., 50 figs., 39 tabs

  15. Structural dynamic analysis of a servo-manipulator for the remote handling of the PRIDE process equipment

    International Nuclear Information System (INIS)

    A slave manipulator which handles a payload in highly hazardous hot cell, is designed to have 6-DOF motions such as pitching and rolling motion of a shoulder joint, a elbow joint, and a wrist joint. Structural dynamic analysis of the slave manipulator needs to be investigated for safe manipulation. In this report, we developed analysis models based on flexible multi-bodies dynamics and performed simulations for some operation cases with predefined tracking trajectories in order to obtain dynamic stress of structures, joint reaction torques, and tensions of cables. The main results are as follows: (1) joint reaction torques, the maximum stress of structures, and cable tensions at dynamic movement are much larger than those of static ones due to the inertia force of payloads which are accelerated to 3g. (2) we could reduce the reaction torques significantly through adjustment of the direction of the inertia force during operation. The results obtained in this study will help the safe manipulation of the slave manipulator, and will be applied to the re-design of the slave manipulator

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

    International Nuclear Information System (INIS)

    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. 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.

  18. Pre-conceptual design assessment of DEMO remote maintenance

    International Nuclear Information System (INIS)

    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

  19. Handling and quantifying uncertainty in geological 3D models: A methodological approach based on remote-sensing and field work.

    Science.gov (United States)

    Baumberger, Roland; Wehrens, Philip; Herwegh, Marco

    2013-04-01

    Geological 3D models are always just an approximation of a complex natural situation. This is especially true in regions, where hard underground data (e.g. bore holes, tunnel mappings and seismic data) is lacking. One of the key problems while developing valid geological 3D models is the three-dimensional spatial distribution of geological structures, particularly with increasing distance from the surface. In our study, we investigate the Alpine 3D Deformation of the crystalline rocks of the Aar massif (Haslital valley, Central Switzerland). Deformation in this area is dominated by different sets of large-scale shear zones, which acted under both ductile and brittle deformation conditions. The goal of our study is the prediction of the geometry and the evolution of the structures in 3D space and time. A key point in our project is the generation of a reliable 3D model of today's structures. In this sense, estimation of the reliability of the surface information for the extrapolation to depth is mandatory. Based on our data, a method will be presented that contributes to a possible solution of the questions addressed above. The basic idea consists of the fact that (i) mechanical anisotropies as shear zones and faults show prominent three-dimensional information in the landscape, (ii) these geometries can be used as input data for a geological 3D model and (iii) that the 3D information mentioned allows a projection to depth. As a great advantage of the study area, a large number of underground tunnels exist, which allow to evaluate the quality of the aforementioned extrapolations. The method is based on a combined remote-sensing and field work approach: morphological incisions recognized on digital elevation models as well as on aerial photos on the computer screen were evaluated, described and attributed in detail in the field. Our approach is based on a six step workflow: (1) Elaboration of a large-scale structural map of geological structures by means of remote

  20. Handling tongs

    International Nuclear Information System (INIS)

    The design is presented of remotely controlled handling tongs for placing fuel assemblies of a fast nuclear reactor in the desired positions in the reactor vessel. The tongs consist of a head and clamps pivoted an the head. The head machined at the end of an inner pull rod which is swing connected to the main pull rod guide bar. The connection is effected from the inner pull rod side. Grip pins are pivoted on the main pull rod guide bar. The side projections of the grip pins engage the inner wall of the channel while the grip pin bodies lean against the opening link. The link pull rod and its height is adjustable. Its inner cut-outs engage the upper tips of the clamps. A fixing ring which the grip pin bodies engage is attached to the opening link such that it can be deflected to both sides. (E.S.)

  1. A method for enabling real-time structural deformation in remote handling control system by utilizing offline simulation results and 3D model morphing

    International Nuclear Information System (INIS)

    A full scale physical test facility, DTP2 (Divertor Test Platform 2) has been established in Finland for demonstrating and refining the Remote Handling (RH) equipment designs for ITER. The first prototype RH equipment at DTP2 is the Cassette Multifunctional Mover (CMM) equipped with Second Cassette End Effector (SCEE) delivered to DTP2 in October 2008. The purpose is to prove that CMM/SCEE prototype can be used successfully for the 2nd cassette RH operations. At the end of F4E grant 'DTP2 test facility operation and upgrade preparation', the RH operations of the 2nd cassette were successfully demonstrated to the representatives of Fusion For Energy (F4E). Due to its design, the CMM/SCEE robot has relatively large mechanical flexibilities when the robot carries the nine-ton-weighting 2nd Cassette on the 3.6-m long lever. This leads into a poor absolute accuracy and into the situation where the 3D model, which is used in the control system, does not reflect the actual deformed state of the CMM/SCEE robot. To improve the accuracy, the new method has been developed in order to handle the flexibilities within the control system's virtual environment. The effect of the load on the CMM/SCEE has been measured and minimized in the load compensation model, which is implemented in the control system software. The proposed method accounts for the structural deformations of the robot in the control system through the 3D model morphing by utilizing the finite element method (FEM) analysis for morph targets. This resulted in a considerable improvement of the CMM/SCEE absolute accuracy and the adequacy of the 3D model, which is crucially important in the RH applications, where the visual information of the controlled device in the surrounding environment is limited.

  2. Realtime graphics support for remote handling operations in complex working environments within the framework of a control, simulation and off-line programming system

    International Nuclear Information System (INIS)

    The application independent simulation system KISMET was developed. This tool gives a different approach compared to previously existing robot simulators. A hierarchical data structure approach is used for the definition of workcell geometry, assembly topology and mechanism kinematics. This database structure allows for presentation of interactively selectable levels of detail and is, therefore, especially useful for real-time rigid body simulation of complex RH-scenarios. With KISMET, assembly structures can be modelled in any number of detail levels. Workcell geometry, assembly topology and mechanisms can be defined interactively by means of the integrated modeller. The mechanism simulation allows for kinematical tree structures with any number of joints, planar closed chains, and interconnections between joints. Examples of novel simulation methods, data structures, and algorithms are presented for selected examples: the hidden surface problem, graphical presentation techniques, collision testing, and control of scene cameras (image simulation, fast positioning and tracking). Special attention is paid to the real-time problem. The way this system was realized within the UNIX world is shown as an example for geometric and kinematic modelling techniques that grant for the optimum use of the capabilities of high-performance graphics workstations. A further chapter is focussing on the use of standard interfaces for CAD model transfer (CAD*I, STEP) and robot programming (IRDATA). Examples of practical KISMET applications for remote handling in fusion reactors, in a nuclear fuel element reprocessing cell and in sensor based robotics are used to present the developed methods. (orig.)

  3. Seismic Characterization of Basalt Topography at Two Candidate Sites for the INL Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Jeff Sondrup; Gail Heath; Trent Armstrong; Annette Shafer; Jesse Bennett; Clark Scott

    2011-04-01

    This report presents the seismic refraction results from the depth to bed rock surveys for two areas being considered for the Remote-Handled Low-Level Waste (RH-LLW) disposal facility at the Idaho National Laboratory. The first area (Site 5) surveyed is located southwest of the Advanced Test Reactor Complex and the second (Site 34) is located west of Lincoln Boulevard near the southwest corner of the Idaho Nuclear Technology and Engineering Center (INTEC). At Site 5, large area and smaller-scale detailed surveys were performed. At Site 34, a large area survey was performed. The purpose of the surveys was to define the topography of the interface between the surficial alluvium and underlying basalt. Seismic data were first collected and processed using seismic refraction tomographic inversion. Three-dimensional images for both sites were rendered from the data to image the depth and velocities of the subsurface layers. Based on the interpreted top of basalt data at Site 5, a more detailed survey was conducted to refine depth to basalt. This report briefly covers relevant issues in the collection, processing and inversion of the seismic refraction data and in the imaging process. Included are the parameters for inversion and result rendering and visualization such as the inclusion of physical features. Results from the processing effort presented in this report include fence diagrams of the earth model, for the large area surveys and iso-velocity surfaces and cross sections from the detailed survey.

  4. Tailorable Advanced Blanket Insulation (TABI)

    Science.gov (United States)

    Sawko, Paul M.; Goldstein, Howard E.

    1987-01-01

    Single layer and multilayer insulating blankets for high-temperature service fabricated without sewing. TABI woven fabric made of aluminoborosilicate. Triangular-cross-section flutes of core filled with silica batting. Flexible blanket formed into curved shapes, providing high-temperature and high-heat-flux insulation.

  5. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    The blanket of the present invention can keep the temperature of breeding materials within a predetermined range even if the breeding materials are consumed and the amount of heat generated from the breeding materials is reduced, thereby enabling to release tritium stably. That is, a neutron incident amount control means is disposed to the blanket for controlling the amount of neutrons incident to the breeding materials. Alternatively, a material to form hollow layers are disposed to the periphery of the breeding materials. With such constitution, the neutron incident amount control means enables to control the incident amount of neutrons from plasmas to the breeding materials, thereby enabling to suppress the change of the amount of heat generated in the breeding materials. In addition, the hollow layers formed at the periphery of the breeding materials enables selective filling of fluids having different heat transfer characteristics thereby enabling to control heat resistance between the breeding materials and cooling tubes. Accordingly, temperature of the breeding materials can be kept constant even in any of the cases. (I.S.)

  6. Design and Preliminary Monte Carlo Calculations of an Active Compton Suppressed LaBr3(Ce) Detector System for TRU Assay in Remote-Handled Wastes

    International Nuclear Information System (INIS)

    Recent studies indicate LaBr3(Ce) scintillation detectors have desirable attributes, such as room temperature operability, which may make them viable alternatives as primary detectors (PD) in a Compton suppression spectrometer (CSS) used for remote-handled transuranic (RH-TRU) waste assay. A CSS with a LaBr3(Ce) PD has been designed and its expected performance evaluated using Monte Carlo analysis. The unique design of this unit minimizes the amount of ''dead'' material between the PD and the secondary guard detector. The analysis results indicate that this detector will have a relatively high Compton-suppression capability, with greater suppression ability for large angle-scattered photons in the PD. J. K. Hartwell1, M. E. McIlwain1, R. P. Gardner2, J. Kulisek3 (1) Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-2114 USA (2) North Carolina State University, Dept of Nuclear Eng., PO Box 7909, Raleigh, NC 27695 USA (3) Ohio State University, Columbus, Ohio 43210 The US Department of Energy's transuranic (TRU) waste inventory includes about 4,500 m3 of remote-handled TRU (RH-TRU) wastes. The RH-TRU waste stream is composed of a variety of containerized waste forms having a contact surface dose rate that exceeds 2 mSv/hr (200 mrem/hr) containing waste materials with a total TRU concentration greater than 3700 Bq/g (100 nCi/g). As part of a research project to investigate the use of active Compton-suppressed room-temperature gamma-ray detectors for direct non-destructive quantification of the TRU content of these RH-TRU wastes, we have designed and purchased a unique detector system using a LaBr3(Ce) primary detector and a NaI(Tl) suppression mantle. The expected detector performance has been modeled using MCNP-X [1] and CEARCPG [2], and incorporates certain design features modeled as important to active Compton suppression systems in previously-published work [3]. The unique detector system is sketched in Fig. 1. The ∼25 mm diameter by 75 mm long LaBr3

  7. Summary of Conceptual Models and Data Needs to Support the INL Remote-Handled Low-Level Waste Disposal Facility Performance Assessment and Composite Analysis

    International Nuclear Information System (INIS)

    An overview of the technical approach and data required to support development of the performance assessment, and composite analysis are presented for the remote handled low-level waste disposal facility on-site alternative being considered at Idaho National Laboratory. Previous analyses and available data that meet requirements are identified and discussed. Outstanding data and analysis needs are also identified and summarized. The on-site disposal facility is being evaluated in anticipation of the closure of the Radioactive Waste Management Complex at the INL. An assessment of facility performance and of the composite performance are required to meet the Department of Energy's Low-Level Waste requirements (DOE Order 435.1, 2001) which stipulate that operation and closure of the disposal facility will be managed in a manner that is protective of worker and public health and safety, and the environment. The corresponding established procedures to ensure these protections are contained in DOE Manual 435.1-1, Radioactive Waste Management Manual (DOE M 435.1-1 2001). Requirements include assessment of (1) all-exposure pathways, (2) air pathway, (3) radon, and (4) groundwater pathway doses. Doses are computed from radionuclide concentrations in the environment. The performance assessment and composite analysis are being prepared to assess compliance with performance objectives and to establish limits on concentrations and inventories of radionuclides at the facility and to support specification of design, construction, operation and closure requirements. Technical objectives of the PA and CA are primarily accomplished through the development of an establish inventory, and through the use of predictive environmental transport models implementing an overarching conceptual framework. This document reviews the conceptual model, inherent assumptions, and data required to implement the conceptual model in a numerical framework. Available site-specific data and data sources

  8. Summary of Conceptual Models and Data Needs to Support the INL Remote-Handled Low-Level Waste Disposal Facility Performance Assessment and Composite Analysis

    Energy Technology Data Exchange (ETDEWEB)

    A. Jeff Sondrup; Annette L. Schafter; Arthur S. Rood

    2010-09-01

    An overview of the technical approach and data required to support development of the performance assessment, and composite analysis are presented for the remote handled low-level waste disposal facility on-site alternative being considered at Idaho National Laboratory. Previous analyses and available data that meet requirements are identified and discussed. Outstanding data and analysis needs are also identified and summarized. The on-site disposal facility is being evaluated in anticipation of the closure of the Radioactive Waste Management Complex at the INL. An assessment of facility performance and of the composite performance are required to meet the Department of Energy’s Low-Level Waste requirements (DOE Order 435.1, 2001) which stipulate that operation and closure of the disposal facility will be managed in a manner that is protective of worker and public health and safety, and the environment. The corresponding established procedures to ensure these protections are contained in DOE Manual 435.1-1, Radioactive Waste Management Manual (DOE M 435.1-1 2001). Requirements include assessment of (1) all-exposure pathways, (2) air pathway, (3) radon, and (4) groundwater pathway doses. Doses are computed from radionuclide concentrations in the environment. The performance assessment and composite analysis are being prepared to assess compliance with performance objectives and to establish limits on concentrations and inventories of radionuclides at the facility and to support specification of design, construction, operation and closure requirements. Technical objectives of the PA and CA are primarily accomplished through the development of an establish inventory, and through the use of predictive environmental transport models implementing an overarching conceptual framework. This document reviews the conceptual model, inherent assumptions, and data required to implement the conceptual model in a numerical framework. Available site-specific data and data sources

  9. Development of the remote-handled transuranic waste radioassay data quality objectives. An evaluation of RH-TRU waste inventories, characteristics, radioassay methods and capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Meeks, A.M.; Chapman, J.A.

    1997-09-01

    The Waste Isolation Pilot Plant will accept remote-handled transuranic waste as early as October of 2001. Several tasks must be accomplished to meet this schedule, one of which is the development of Data Quality Objectives (DQOs) and corresponding Quality Assurance Objectives (QAOs) for the assay of radioisotopes in RH-TRU waste. Oak Ridge National Laboratory (ORNL) was assigned the task of providing to the DOE QAO, information necessary to aide in the development of DQOs for the radioassay of RH-TRU waste. Consistent with the DQO process, information needed and presented in this report includes: identification of RH-TRU generator site radionuclide data that may have potential significance to the performance of the WIPP repository or transportation requirements; evaluation of existing methods to measure the identified isotopic and quantitative radionuclide data; evaluation of existing data as a function of site waste streams using documented site information on fuel burnup, radioisotope processing and reprocessing, special research and development activities, measurement collection efforts, and acceptable knowledge; and the current status of technologies and capabilities at site facilities for the identification and assay of radionuclides in RH-TRU waste streams. This report is intended to provide guidance in developing the RH-TRU waste radioassay DQOs, first by establishing a baseline from which to work, second, by identifying needs to fill in the gaps between what is known and achievable today and that which will be required before DQOs can be formulated, and third, by recommending measures that should be taken to assure that the DQOs in fact balance risk and cost with an achievable degree of certainty.

  10. 324 Building Compliance Project: Selection and evaluation of alternatives for the removal of solid remote-handled mixed wastes from the 324 Building

    International Nuclear Information System (INIS)

    Six alternatives for the interim storage of remote-handled mixed wastes from the 324 Building on the Hanford Site have been identified and evaluated. The alternatives focus on the interim storage facility and include use of existing facilities in the 200 Area, the construction of new facilities, and the vitrification of the wastes within the 324 Building to remove the majority of the wastes from under RCRA regulations. The six alternatives are summarized in Table S.1, which identifies the primary facilities to be utilized, the anticipated schedule for removal of the wastes, the costs of the transfer from 324 Building to the interim storage facility (including any capital costs), and an initial risk comparison of the alternatives. A recently negotiated Tri-Party Agreement (TPA) change requires the last of the mixed wastes to be removed by May 1999. The ability to use an existing facility reduces the costs since it eliminates the need for new capital construction. The basic regulatory approvals for the storage of mixed wastes are in place for the PUREX facility, but the Form HI permit will need some minor modifications since the 324 Building wastes have some additional characteristic waste codes and the current permit limits storage of wastes to those from the facility itself. Regulatory reviews have indicated that it will be best to use the tunnels to store the wastes. The PUREX alternatives will only provide storage for about 65% of the wastes. This results from the current schedule of the B-Cell Clean Out Project, which projects that dispersible debris will continue to be collected in small quantities until the year 2000. The remaining fraction of the wastes will then be stored in another facility. Central Waste Complex (CWC) is currently proposed for that residual waste storage; however, other options may also be available

  11. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    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

  12. Blanket comparison and selection study. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    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. (MOW)

  13. First wall and blanket module safety enhancement by material selection and design decision

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J.

    1980-01-01

    A thermal/mechanical study has been performed which illustrates the behavior of a fusion reactor first wall and blanket module during a loss of coolant flow event. The relative safety advantages of various material and design options were determined. A generalized first wall-blanket concept was developed to provide the flexibility to vary the structural material (stainless steel vs titanium), coolant (helium vs water), and breeder material (liquid lithium vs solid lithium aluminate). In addition, independent vs common first wall-blanket cooling and coupled adjacent module cooling design options were included in the study. The comparative analyses were performed using a modified thermal analysis code to handle phase change problems.

  14. Evaluation of Groundwater Impacts to Support the National Environmental Policy Act Environmental Assessment for the INL Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Annette Schafer, Arthur S. Rood, A. Jeffrey Sondrup

    2011-12-23

    Groundwater impacts have been analyzed for the proposed remote-handled low-level waste disposal facility. The analysis was prepared to support the National Environmental Policy Act environmental assessment for the top two ranked sites for the proposed disposal facility. A four-phase screening and analysis approach was documented and applied. Phase I screening was site independent and applied a radionuclide half-life cut-off of 1 year. Phase II screening applied the National Council on Radiation Protection analysis approach and was site independent. Phase III screening used a simplified transport model and site-specific geologic and hydrologic parameters. Phase III neglected the infiltration-reducing engineered cover, the sorption influence of the vault system, dispersion in the vadose zone, vertical dispersion in the aquifer, and the release of radionuclides from specific waste forms. These conservatisms were relaxed in the Phase IV analysis which used a different model with more realistic parameters and assumptions. Phase I screening eliminated 143 of the 246 radionuclides in the inventory from further consideration because each had a half-life less than 1 year. An additional 13 were removed because there was no ingestion dose coefficient available. Of the 90 radionuclides carried forward from Phase I, 57 radionuclides had simulated Phase II screening doses exceeding 0.4 mrem/year. Phase III and IV screening compared the maximum predicted radionuclide concentration in the aquifer to maximum contaminant levels. Of the 57 radionuclides carried forward from Phase II, six radionuclides were identified in Phase III as having simulated future aquifer concentrations exceeding maximum contaminant limits. An additional seven radionuclides had simulated Phase III groundwater concentrations exceeding 1/100th of their respective maximum contaminant levels and were also retained for Phase IV analysis. The Phase IV analysis predicted that none of the thirteen remaining

  15. Issues and Recommendations Arising from the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility Composite Analysis - 13374

    International Nuclear Information System (INIS)

    Development of the composite analysis (CA) for the Idaho National Laboratory's (INLs) proposed remote-handled (RH) low-level waste (LLW) disposal facility has underscored the importance of consistency between analyses conducted for site-specific performance assessments (PAs) for LLW disposal facilities, sites regulated by the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) [1], and residual decontamination and decommissioning (D and D) inventories. Consistency is difficult to achieve because: 1) different legacy sources and compliance time-periods were deemed important for each of the sites evaluated at INL (e.g., 100 years for CERCLA regulated facilities vs. 1,000 years for LLW disposal facilities regulated under U.S. Department of Energy (DOE) Order 435.1 [2]); 2) fate and transport assumptions, parameters, and models have evolved through time at the INL including the use of screening-level parameters vs. site-specific values; and 3) evaluation objectives for the various CERCLA sites were inconsistent with those relevant to either the PA or CA including the assessment of risk rather than effective dose. The proposed single site-wide CA approach would provide needed consistency, allowing ready incorporation of new information and/or facilities in addition to being cost effective in terms of preparation of CAs and review by the DOE. A single site-wide CA would include a central database of all existing INL sources, including those from currently operating LLW facilities, D and D activities, and those from the sites evaluated under CERCLA. The framework presented for the INL RH-LLW disposal facility allows for development of a single CA encompassing air and groundwater impacts. For groundwater impacts, a site-wide MODFLOW/MT3D-MS model was used to develop unit-response functions for all potential sources providing responses for a grid of receptors. Convolution and superposition of the response functions are used to compute groundwater

  16. Activation analyses for the different options considered in the US ITER blanket trade-off study

    International Nuclear Information System (INIS)

    Detailed activation analyses were performed for the different blanket design options considered in the ITER blanket option trade-off study. The options considered included a self-cooled Li/V option, a helium-cooled Li/V option and a water-cooled 316 SS non-breeding shield option. A vacuum vessel made of double-wall Inconel 625 and water-cooled 316 SS balls was used with all options. The He-cooled blanket activity is higher than that of the self-cooled blanket due to the larger structure content. Meanwhile, the vacuum vessel activity is lower for the He-cooled blanket option due to the larger neutron attenuation in the blanket. The shield activity and decay heat of the 316 SS/H2O option are higher than those of the Li/V blankets due to the large amount (80%) of 316 SS used. In both Li/V options the blanket qualifies as class C low-level waste. On the other hand, the 316 SS/H2O shield does not qualify for disposal as low- level waste. The 316 SS/H2O option produces the highest off-site doses in the case of accidental release of 100% of its radioactive inventory. Only remote maintenance would be allowed for all options. (orig.)

  17. Current status of safety design and safety analysis for China ITER helium coolant ceramic breeder test blanket system long

    International Nuclear Information System (INIS)

    Helium Coolant Ceramic Breeder (HCCB) Test Blanket System (TBS) designed by China are planned to be tested in ITER to validate key technologies, including demonstration of nuclear safety, for future fusion reactor breeding blankets. Furthermore, in order to be operated in ITER, a nuclear facility (INB) recognized by French nuclear safety authority, safety design and safety analysis of the TBS are mandatory for the licensing procedures. This paper summarizes the status at current design phase with following main elements: The main radiological source terms in the system are tritium and activation products. Nuclear and tritium analysis are performed to identify their inventories and distributions in system. Multiple confinement barriers are considered to be the most essential safety feature. French regulation for pressure equipment and nuclear equipment (ESP/ESPN regulations) will be followed to ensure the system integrities. ALARA principle is kept in mind during the whole safety design phases. Protective actions including choice of advanced materials, improvement of shielding, optimization of operation and maintenance activities, usage of remote handling operations, zoning and access control have been considered. Passive safety is emphasized in the system design, only minimal active safety functions including call for fusion plasma shutdown and isolation of TBM from ex-vessel ancillary systems. High reliability and redundancies are required for components related to these functions. Several accidents have been identified and analyzed. Consider the limited inventories in the system and the intrinsic safety of fusion device, positive conclusions have been obtained. (author)

  18. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    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

  19. Assessment of Potential Flood Events and Impacts at INL's Proposed Remote-Handled Low-Level Waste Disposal Facility Sites

    Energy Technology Data Exchange (ETDEWEB)

    A. Jeff Sondrup; Annette L. Schafter

    2010-09-01

    Rates, depths, erosion potential, increased subsurface transport rates, and annual exceedance probability for potential flooding scenarios have been evaluated for the on-site alternatives of Idaho National Laboratory’s proposed remote handled low-level waste disposal facility. The on-site disposal facility is being evaluated in anticipation of the closure of the Radioactive Waste Management Complex at the INL. An assessment of flood impacts are required to meet the Department of Energy’s Low-Level Waste requirements (DOE-O 435.1), its natural phenomena hazards assessment criteria (DOE-STD-1023-95), and the Radioactive Waste Management Manual (DOE M 435.1-1) guidance in addition to being required by the National Environmental Policy Act (NEPA) environmental assessment (EA). Potential sources of water evaluated include those arising from (1) local precipitation events, (2) precipitation events occurring off of the INL (off-site precipitation), and (3) increased flows in the Big Lost River in the event of a Mackay Dam failure. On-site precipitation events include potential snow-melt and rainfall. Extreme rainfall events were evaluated for the potential to create local erosion, particularly of the barrier placed over the disposal facility. Off-site precipitation carried onto the INL by the Big Lost River channel was evaluated for overland migration of water away from the river channel. Off-site precipitation sources evaluated were those occurring in the drainage basin above Mackay Reservoir. In the worst-case scenarios, precipitation occurring above Mackay Dam could exceed the dam’s capacity, leading to overtopping, and eventually complete dam failure. Mackay Dam could also fail during a seismic event or as a result of mechanical piping. Some of the water released during dam failure, and contributing precipitation, has the potential of being carried onto the INL in the Big Lost River channel. Resulting overland flows from these flood sources were evaluated for

  20. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    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

  1. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    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 usable 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

  2. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    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

  3. Comparison analysis of fusion breeder blanket concepts

    International Nuclear Information System (INIS)

    Based on the wide survey, the development status and key issues of fusion breeder blanket concepts are summarized. Two types of blanket concepts, i.e. solid and liquid breeder blanket, were compared and assessed in terms of engineering feasibility, tritium recovery and control, economic and safety aspects, etc. The advantages and disadvantages of the two types of blanket concepts were clarified from the viewpoint of technology realization and development potential. This study may act as a valuable reference for fusion blanket concept selection and design. (authors)

  4. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    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)

  5. Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

    International Nuclear Information System (INIS)

    In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained

  6. Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained.

  7. Development of a handling system for the remote controlled dismantling of the steamdryer housing of KRB/A Gundremmingen -ODIN 1

    International Nuclear Information System (INIS)

    As the first generation of civil nuclear power plants reaches the end of their service life, decommissioning of these facilities becomes more and more important to the highly industrialized nations, opening a wide field for investigations on dismantling and handling techniques. Several nuclear installations are being dismantled in Europe and all over the world. One of these is Block A of KRB Gundremminger in Germany. The first component from the reactor core to be segmented is the steamdryer housing of DRB A. It will be cut in the deplaning pool of the reactor, i.e. under water, with the use of plasma arc and consumable electrode waterjet, both thermal cutting techniques. With this dismantling task, experience and know how will be gained concerning cutting and handling techniques, especially in nuclear environments. This paper outlines the dismantling task and describes briefly the design of the tool guiding device ODIN I, which was developed at the Institut fur Werkstoffkunde, University of Hanover, Germany, for this particular cutting problem. (Author)

  8. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    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

  9. Technical evaluation of major candidate blanket systems for fusion power reactor

    International Nuclear Information System (INIS)

    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/Li2O/H2O/Be, Mo-alloy/Li2O/He/Be, Mo-alloy/LiAlO2/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)

  10. Compatibility problems in tritium breeding blankets

    International Nuclear Information System (INIS)

    Compatibility between tritium breeding materials (liquid or solid), neutron multiplier and structural steels is a concern for the choice of a tritium breeding blanket for NET. For solid tritium breeding blanket, it seems that the more severe compatibility problem is due to the interaction of beryllium with steel. As for the water-cooled Pb17Li blanket, the first results obtained in experimental conditions closed to the concept have evidenced lower corrosion rates than those measured in thermal convection loops

  11. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

    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

  12. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    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

  13. Fusion reactor blanket/shield design study

    International Nuclear Information System (INIS)

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented

  14. ITER cryopump handling study preliminary report

    International Nuclear Information System (INIS)

    The scope of this study was to develop a remote handling concept for the primary vacuum pumping system by performing the following tasks: review the cryopump duct configuration and remote handling requirements for cryopump replacement; initiate an ITER Cryopump Maintenance Requirements Definition Document; develop a preliminary, compatible design concept ; propose preliminary maintenance procedures for cryopump replacement; develop a preliminary development schedule, breakdown structure and ROM costs. 3 tabs., 4 figs

  15. ITER blanket, shield and material data base

    International Nuclear Information System (INIS)

    As part of the summary of the Conceptual Design Activities (CDA) for the International Thermonuclear Experimental Reactor (ITER), this document describes the ITER blanket, shield, and material data base. Part A, ''ITER Blanket and Shield Conceptual Design'', discusses the need for ITER of a tritium breeding blanket to supply most of the tritium for the fuel cycle of the device. Blanket and shield combined must be designed to operate at a neutron wall loading of 1MW/m2, and to provide adequate shielding of the magnets to meet the neutron energy fluence goal of 3MWa/m2 at the first wall. After a summary of the conceptual design, the following topics are elaborated upon: (1) function, design requirement, and critical issues; (2) material selection; (3) blanket and shield segmentation; (4) blanket design description; (5) design analysis; (6) shield; (7) radiation streaming analysis; and (8) a summary of benchmark calculations. Part B, ''ITER Materials Evaluation and Data Base'', treats the compilation and assessment of the available materials data base used for the selection of the appropriate materials for all major components of ITER, including (i) structural materials for the first wall, (ii) Tritium breeding materials for the blanket, (iii) plasma facing materials for the divertor and first wall armor, and (4) electric insulators for use in the blanket and divertor. Refs, figs and tabs

  16. Lithium-cooled blankets for advanced tokamaks

    International Nuclear Information System (INIS)

    The main objective of the Tokamak Power System Studies (TPSS) at Argonne National Lab. during fiscal year 1985 was to explore innovative design concepts that have the potential for significant enhancement of the attractiveness of a tokamak-based power plant. Activities in the area of plasma engineering resulted in a reference reactor concept, which served as a model for the impurity control and first-wall/blanket/shield studies. The liquid-metal-cooled first-wall/blanket/shield design activity was centered around the vanadium alloy structure and liquid-lithium coolant leading blanket concept as identified by the Blanket Comparison and Selection Study (BCSS). A ferritic steel structure and a LiPb breeder were considered as backup options. The magnetohydrodynamics (MHD) effects associated with self-cooled liquid-metal blanket/first-wall systems are substantially reduced by the lower magnetic fields required for higher plasmas, the lower neutron wall loading resulting from reduced power output, and the smaller reactor size of the TPSS model reactor. Therefore, improved performance characteristics of self-cooled liquid-metal blanket concepts are achievable mainly because the design constraints are more relaxed compared to the BCSS guidelines. Key aspects of the designs evaluated in the current study include the following: (1) design simplicity; (2) use of the first wall as an impurity control device; (3) modular first-wall/blanket/reflector/shield construction; and (4) integrated first-wall/blanket/reflector/shield

  17. Design and analysis of ITER shield blanket

    International Nuclear Information System (INIS)

    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.)

  18. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    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)

  19. Disruption problematics in segmented-blanket concepts

    International Nuclear Information System (INIS)

    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 segmented-blanket design concepts in order to satisfy the disruption problematics.The present paper describes recent investigations on internal blanket reinforcement systems needed in order to improve the first-wall/blanket/shield structural design for next-step and commercial fusion reactors. Particularly in the context of SEAFP and ITER activities, representative 3-D CAD models of the inboard and outboard blanket regions and the related magnetomechanical simulations are illustrated. (orig.)

  20. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    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....

  1. Development of Solid Breeder Blanket at JAERI

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been performing blanket development based on the long-term research program of fusion blankets in Japan, which was approved by the Fusion Council of Japan in 1999. The blanket development consists of out-pile R and D, In-pile R and D, TBM Neutronics and TPR Tests and Tritium Recovery System R and D. Based on the achievements of element technology development, the R and D program is now stepping to the engineering testing phase, in which scalable mockup tests will be performed for obtaining engineering data unique to the specific structure of the components, with the objective to define the fabrication specification of test blanket modules for ITER. This paper presents the major achievements of the element technology development of solid breeder blanket in JAERI

  2. Remote handling in the Plutonium Immobilization Project: Puck handling

    International Nuclear Information System (INIS)

    Since the break up of the Soviet Union at the end of the Cold War, the US and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

  3. Puck Handling Glovebox

    International Nuclear Information System (INIS)

    The Plutonium Immobilization Project (PIP) is a joint venture between the Savannah River Site (SRS) and Lawrence Livermore National Laboratory (LLNL). This project will disposition excess weapons grade plutonium in a solid ceramic form. The plutonium, in oxide powder form, will be mixed with uranium oxide powder, ceramic precursors and binders. The combined powder mixture will be milled and possibly granulated; this processed powder will then be dispensed to a (dual action) cold press where it will be formed into green (unsintered) compacts. The compact will have the shape of a puck measuring approximately 3 1/2'' in diameter and 1 3/8'' thick. The green puck, once ejected from the press die, will be picked up by a robot and transferred into the Puck Handling Glovebox. Here the green puck will be inspected and then palletized onto furnace trays. The loaded furnace trays will be stacked/assembled and transported to the furnace where sintering operations will be performed. Finally the sintered pucks will be off loaded, inspected and transferred onto Transfer Trays which will carry the pucks from the Puck Handling Glovebox downstream to subsequent Bagless Transfer Can (BTC) operations. Due to contamination potential and high radiation rates, all Puck Handling Glovebox operations will be performed remotely using robots and specialized automation

  4. Fast breeder reactor blanket management: comparison of LMFBR and GCFR blankets

    International Nuclear Information System (INIS)

    The economic performance of the fast breeder reactor blanket, considering different fuel management schemes was studied. To perform this, the investigation started with a standard reactor physics calculation. Then, two economic models for evaluation of the economic performance of the radial blanket were developed. These models formed the basis of a computer code, ECOBLAN, which computes the net economic gain and the levelized fuel cost due to the radial blanket. The net gain in terms of dollars and $/kgHM-y and the levelized fuel cost in mills/kWhe were obtained as a function of blanket thickness and a residence time of the fuel in the blanket. A LMFBR and a GCFR were the reactor models considered in this study. The optimum radial blanket of a GCFR consists of two rows, that of a LMFBR consists of three rows. Regarding the different fuel management schemes, the fixed blanket was found to be more favorable than reshuffled blanket. Out-in and in-out reshuffled blanket offer almost the same net gain. In all the cases, the burnup calculated for the fuel was found to be less than the acceptable limit. There is an optimum residence time for the fuel in the blanket which depends on the position of the fuel in the blanket and the fuel management scheme studied. As expected, except for very short residence times (less than 2.5 years), the radial blanket is a net income producer. There is no significant difference between the economic performance of the blanket of a LMFBR and a GCFR

  5. Remote maintenance considerations for swimming pool tokamak reactor

    International Nuclear Information System (INIS)

    Swimming Pool Tokamak Reactor (SPTR) is one of the candidate devices which are expected to demonstrate physical and engineering feasibility for fusion power reactors. In SPTR, water shield is adopted instead of solid shield structures. Among the advantages of SPTR are, from viewpoint of remote maintenance, small handling weight and high space availability between TF coils and a vacuum vessel. On the other hand, high dose rate during reactor repair and adverse effects on remote maintenance equipment by the shielding water might be the disadvantage of SPTR, where it is assumed that the shielding water is drained during reactor repair. Since the design of SPTR is still at the preliminary stage, for remote maintenance, much effort has been directed to clarification of design conditions such as environment and handling weight. As for the remote maintenance system concepts, studies have been focussed on those for a vacuum vessel and its internal structure (blanket, divertor and protection walls) expected to be repaired more frequently. The vacuum vessel assembly is divided into 21 sectors and number of TF coils is 14. A pair of TF coils are connected with each other by antitorque beams on the whole side surface. Vacuum vessel cassettes and associated blanket, divertor and protection walls are replaced through seven windows between TF coils pairs. Therefore each vacuum vessel cassette is required moving mechanisms in toroidal and radial directions. Options for slide mechanisms are wheels, balls, rollers and water bearings. Options for driving the cassette are self-driving by hydraulic motors and external driving by rack-pinion, wires or specific vehicles. As a result of studies, the moving mechanism with wheels and hydraulic motors has been selected for the reference design, and the system with water bearings and rack-pinion as an alternative. Furthermore typical concepts have been obtained for remote maintenance equipment such as wall-mounted manipulators, tools for

  6. ITER reference breeding blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Ferrari, M. [ENEA, Frascati (Italy); Bianchi, A. [EFET, Ansaldo Ricerche, Genova (Italy); Celentano, G. [ENEA, ERG-FUS, Centro di Frascati, Via Enrico Fermi, 27, P.O. Box 65, I-00044, Frascati (IT)] [and others

    1999-11-01

    The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions. (orig.)

  7. ITER reference breeding blanket design

    International Nuclear Information System (INIS)

    The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions. (orig.)

  8. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

    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

  9. Multivariable optimization of fusion reactor blankets

    International Nuclear Information System (INIS)

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% 6Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO2 breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO2 breeding blanket enriched to 34% 6Li

  10. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    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 m2) 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.)

  11. 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)

  12. ITER breeding blanket module design and analysis

    International Nuclear Information System (INIS)

    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)

  13. Preparing, Loading and Shipping Irradiated Metals in Canisters Classified as Remote-Handled (RH) Low-Level Waste (LLW) From Oak Ridge National Laboratory (ORNL) to the Nevada Test Site (NTS)

    International Nuclear Information System (INIS)

    Irradiated metals, classified as remote-handled low-level waste generated at the Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee, were containerised in various sized canisters for long-term storage. The legacy waste canisters were placed in below-grade wells located at the 7827 Facility until a pathway for final disposal at the Nevada Test Site (NTS) could be identified and approved. Once the pathway was approved, WESKEM, LLC was selected by Bechtel Jacobs Company, LLC to prepare, load, and ship these canisters from ORNL to the NTS. This paper details some of the technical challenges encountered during the retrieval process and solutions implemented to ensure the waste was safely and efficiently over-packed and shipped for final disposal. The technical challenges detailed in this paper include: 1) how to best perform canister/lanyard pre-lift inspections since some canisters had not been moved in ∼10 years, so deterioration was a concern; 2) replacing or removing damaged canister lanyards; 3) correcting a mis-cut waste canister lanyard resulting in a shielded overpack lid not seating properly; 4) retrieving a stuck canister; and 5) developing a path forward after an overstrained lanyard failed causing a well shield plug to fall and come in contact with a waste canister. Several of these methods can serve as positive lessons learned for other projects encountering similar situations. (authors)

  14. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    The thermal and mechanical aspects of hybrid reactor blanket design considerations are discussed. This paper is intended as a companion to that of J. D. Lee of Lawrence Livermore Laboratory on the nuclear aspects of hybrid reactor blanket design. The major design characteristics of hybrid reactor blankets are discussed with emphasis on the areas of difference between hybrid reactors and standard fusion or fission reactors. Specific examples are used to illustrate the design tradeoffs and choices that must be made in hybrid reactor design. These examples are drawn from the work on the Mirror Hybrid Reactor

  15. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li17Pb83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li17Pb83 blankets. (author)

  16. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

    Comparisons of alternative fusion reactor blanket/coolant systems suggest that environmental considerations will enter strongly into selection of design and materials. Liquid blankets and coolants tend to maximize transport of radioactive corrosion products. Liquid lithium interacts strongly with tritium, minimizing permeation and escape of gaseous tritium in accidents. However, liquid lithium coolants tend to create large tritium inventories and have a large fire potential compared to flibe and solid blankets. Helium coolants minimize radiation transport, but do not have ability to bind the tritium in case of accidental releases. (auth)

  17. Divertor and gas blanket impurity control study

    International Nuclear Information System (INIS)

    A simple calculational model for the transport of particles across the scrap off region between the plasma and the wall in the presence of a divertor or a gas blanket has been developed. The model departs from previous work in including: (a) the entire impurity transport as well as its effect on the energy balance equations; (b) the recycling neutrals from the divertor, and (c) the reflected neutrals from the wall. Results obtained with this model show how the steady state impurity level in the plasma depends on the divertor parameters such as the neutral backflow from the divertor, the particle residence time and the scrape off thickness; and on the gas blanket parameters such as the neutral source strength and the gas blanket thickness. The variation of the divertor or gas blanket performance as a function of the heat and particle fluxes escaping from the plasma, the wall material and the cross field diffusion is examined and numerical examples are given

  18. Exploratory Study of Blanket Liquid Curtain

    Institute of Scientific and Technical Information of China (English)

    HUGang; HUANGJinhua; FENGKaiming

    2003-01-01

    Blankets and other in-vessel components are easily damaged owing to their circumstance of high radiation and high heat. To protect them, first wall design should be considered. Owing to its high heat removal nd self-refreshing capability, liquid metal first wall has been seen as a potential first wall for a fusion reactor in the future. Blanketliquid curtain is actually a special liquid metal wall to protect blanket.

  19. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

    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

  20. MHD pressure drop at bare welding positions in pipes of DCLL blankets (KIT Scientific Reports ; 7636)

    OpenAIRE

    Bühler, Leo

    2013-01-01

    A systematic parametric analysis has been performed using asymptotic numerical methods for determination of MHD flows near gaps of electrically insulating inserts in well conducting pipes. Such gaps could be present at several positions in fusion blankets, where cutting and rewelding by remotely controlled tools is foreseen. Gaps in the insulation provide additional current paths which leads to increased current density and braking electromagnetic Lorentz forces.

  1. A graphics based remote handling control system

    International Nuclear Information System (INIS)

    A control and simulation system with an interactive graphic man-machine interface is proposed for the articulated boom in JET. The system shall support 1. the study of boom movements in the planning phase, 2. the training of operators by appropriate simulations, 3. the programming of boom movements, and 4. the on-line control of the boom. A combination of computer graphic display and TV-images is proposed for providing optimum recognition of the actual situation and for echoing to the operator actions. (orig.)

  2. Design and testing of a unique active Compton-suppressed LaBr3(Ce) detector system for improved sensitivity assays of TRU in remote-handled TRU wastes

    International Nuclear Information System (INIS)

    The US Department of Energy's transuranic (TRU) waste inventory includes about 4,500 m3 of remote-handled TRU (RH-TRU) wastes composed of a variety of containerized waste forms having a contact surface dose rate that exceeds 2 mSv/hr (200 mrem/hr) containing waste materials with a total TRU concentration greater than 3700 Bq/g (100 nCi/g). As part of a research project to investigate the use of active Compton-suppressed room-temperature gamma-ray detectors for direct non-destructive quantification of the TRU content of these RH-TRU wastes, we have designed and purchased a unique detector system using a LaBr3(Ce) primary detector and a NaI(Tl) suppression mantle. The LaBr3(Ce) primary detector is a cylindrical unit ∼25 mm in diameter by 76 mm long viewed by a 38 mm diameter photomultiplier. The NaI(Tl) suppression mantle (secondary detector) is 175 mm by 175 mm with a center well that accommodates the primary detector. An important feature of this arrangement is the lack of any 'can' between the primary and secondary detectors. These primary and secondary detectors are optically isolated by a thin layer (.003') of aluminized kapton, but the hermetic seal and thus the aluminum can surrounds the outer boundary of the detector system envelope. The hermetic seal at the primary detector PMT is at the PMT wall. This arrangement virtually eliminates the 'dead' material between the primary and secondary detectors, a feature that preliminary modeling indicated would substantially improve the Compton suppression capability of this device. This paper presents both the expected performance of this unit determined from modeling with MCNPX, and the performance measured in our laboratory with radioactive sources

  3. Production enhancement and quality degradation of Pu produced in FBR blankets

    International Nuclear Information System (INIS)

    EOC. The quality of Pu is generally classified into super (30%) garade based on the content of Pu-240 in Pu. By referring this classification, the quality of Pu generated in the blanket is deteriorated from super grade to fuel grade if 10-15% ZrH is loaded into blanket. Side effects caused by ZrH pin loading are also studied in terms of sodium void reactivity and power distribution. The loading of ZrH slightly decrease the sodium void reactivity however the impact is not significant. The power density at EOC in the radial blanket is almost doubled by loading 30% of ZrH compared with no ZrH loading case. The effects of moderator (ZrH) loading into radial blanket on Pu production efficiency and Pu quality deterioration are studied intending for better economy and enhanced proliferation resistance of FBR fuel cycle. The loading ZrH contributes to improve Pu production efficiency however FIR slightly improved just in the low range of ZrH (<5%). The impact of small ZrH loading on sodium void reactivity and power distribution is not significant and those can be handled by core design optimization. (author)

  4. Safety handling of beryllium for fusion technology R and D

    International Nuclear Information System (INIS)

    Feasibility of beryllium use as a blanket neutron multiplier, first wall and plasma facing material has been studied for the D-T burning experiment reactors such as ITER. Various experimental work of beryllium and its compounds will be performed under the conditions of high temperature and high energy particle exposure simulating fusion reactor conditions. Beryllium is known as a hazardous substance and its handling has been carefully controlled by various health and safe guidances and/or regulations in many countries. Japanese regulations for hazardous substance provide various guidelines on beryllium for the protection of industrial workers and environment. This report was prepared for the safe handling of beryllium in a laboratory scale experiments for fusion technology R and D such as blanket development. Major items in this report are; (1) Brief review of guidances and regulations in USA, UK and Japan. (2) Safe handling and administration manuals at beryllium facilities in INEL, LANL and JET. (3) Conceptual design study of beryllium handling facility for small to mid-scale blanket R and D. (4) Data on beryllium toxicity, example of clinical diagnosis of beryllium disease, and environmental occurence of beryllium. (5) Personnel protection tools of Japanese Industrial Standard for hazardous substance. (author) 61 refs

  5. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for test blanket module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation four different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port two TBMs will be accommodated, in the port 16 will be the European helium-cooled pebble bed blanket. In different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for the exchange of TBMs. Two TBMs are mounted onto one common frame, into a port plug (PP), which offers a standardised interface to the vacuum vessel (VV). It is cantilevered with a flange to VV port extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs is also the standard operation of ITER. Several components of the helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system are integrated into the ancillary equipment unit (AEU), which during the operation will connect the port plug to the subsystems. The bigger part of the AEU is accommodated in the port cell and the rest part of it is penetrated into the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection/disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected here. A

  6. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for Test Blanket Module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation 4 different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port will be two TBMs accomodated, in the port 16 will be the the European Helium Cooled Pebble Bed blanket. In the different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for exchange the TBMs. Two TBMs are mounted into one common frame, into a Port Plug (PP), which offers a standardised interface to the Vacuum Vessel (VV). It is cantilevered with a flange to VV Port Extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs are also standard operation of ITER. Several components of the Helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system, etc. All of them is integrated into the Ancillary Equipment Unit (AEU) which during operation will connect the port plug to the sub systems. The bigger part of the AEU is accomodated in the Port Cell and the rest part of it is penetrate to the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection / disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected

  7. Blanket management method for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    A method for reducing thermal striping in liquid metal fast breeder reactors by reducing temperature gradients between adjacent fuel and blanket assemblies by shuffling blanket assemblies at each refueling outage so as to progressively shuffle the blanket assemblies to the core periphery through multiple moves and to generally locate fresh blanket assemblies adjacent to exposed fuel assemblies and exposed blanket assemblies adjacent to fresh fuel. Additionally, assembly orificing is altered to provide less flow to blanket assemblies needing less flow due to an otherwise decreased temperature gradient and providing additional flow to fuel assemblies which need more flow to sufficiently reduce temperature gradients to prevent thermal striping. (author)

  8. Solid breeder blanket design and tritium breeding

    International Nuclear Information System (INIS)

    Thermonuclear D-T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements - to breed one tritium per fusion neutron - are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li2O, LiAlO2, Li2ZrO3, Li4SiO4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the amount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket. (orig.)

  9. SRV-automatic handling device

    International Nuclear Information System (INIS)

    Automatic handling device for the steam relief valves (SRV's) is developed in order to achieve a decrease in exposure of workers, increase in availability factor, improvement in reliability, improvement in safety of operation, and labor saving. A survey is made during a periodical inspection to examine the actual SVR handling operation. An SRV automatic handling device consists of four components: conveyor, armed conveyor, lifting machine, and control/monitoring system. The conveyor is so designed that the existing I-rail installed in the containment vessel can be used without any modification. This is employed for conveying an SRV along the rail. The armed conveyor, designed for a box rail, is used for an SRV installed away from the rail. By using the lifting machine, an SRV installed away from the I-rail is brought to a spot just below the rail so that the SRV can be transferred by the conveyor. The control/monitoring system consists of a control computer, operation panel, TV monitor and annunciator. The SRV handling device is operated by remote control from a control room. A trial equipment is constructed and performance/function testing is carried out using actual SRV's. As a result, is it shown that the SRV handling device requires only two operators to serve satisfactorily. The required time for removal and replacement of one SRV is about 10 minutes. (Nogami, K.)

  10. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. 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. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.)

  11. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    There is a considerable world-wide effort directed toward the production of materials for fusion reactors. Many ceramic fabrication groups are working on making lithium ceramics in a variety of forms, to be incorporated into the tritium breeding blanket which will surround the fusion reactor. Current blanket designs include ceramic in either monolithic or packed sphere bed (sphere-pac) forms. The major thrust at AECL is the production of lithium aluminate spheres to be incorporated in a sphere-pac bed. Contemporary studies on breeder blanket design offer little insight into the requirements on the sizes of the spheres. This study examined the parameters which determine the properties of pressure drop and coolant requirements. It was determined that an optimised sphere-pac bed would be composed of two diameters of spheres: 75 weight % at 3 mm and 25 weight % at 0.3 mm

  12. Neutronic implications of lead-lithium blankets

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W.R.

    1982-08-01

    Lead-lithium alloys have been proposed for use in several conceptual blanket designs for both inertial and magnetic confinement fusion reactors. In most cases, Pb/sub 83/Li/sub 17/, a eutectic with a melting point of 235/sup 0/C, is the chosen composition. The primary reasons for using Pb/sub 83/Li/sub 17/ instead of Li as the tritium breeding material are the perceived safety advantages, low tritium solubility, and favorable neutronic characteristics. This paper describes the neutronic characteristics of Pb/sub 83/Li/sub 17/ blankets with emphasis on the enhanced neutron leakage through chamber ports and the degradation in blanket performance parameters that occurs as a result of the enhanced leakage.

  13. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    The Tandem Mirror Reactor, with its cylindrical central cell, has led to numerous blanket designs taking advantage of the simple geometry. Also many new applications for fusion neutrons are now being considered. To the pure fusion electricity producers and hybrids producing fissile fuel, we are adding studies of synthetic fuel producers and fission-suppressed hybrids. The three blanket concepts presented are new ideas and should be considered illustrative of the breadth of Livermore's application studies. They are not meant to imply fully analyzed designs

  14. Lightweight IMM PV Flexible Blanket Assembly

    Science.gov (United States)

    Spence, Brian

    2015-01-01

    Deployable Space Systems (DSS) has developed an inverted metamorphic multijunction (IMM) photovoltaic (PV) integrated modular blanket assembly (IMBA) that can be rolled or z-folded. This IMM PV IMBA technology enables a revolutionary flexible PV blanket assembly that provides high specific power, exceptional stowed packaging efficiency, and high-voltage operation capability. DSS's technology also accommodates standard third-generation triple junction (ZTJ) PV device technologies to provide significantly improved performance over the current state of the art. This SBIR project demonstrated prototype, flight-like IMM PV IMBA panel assemblies specifically developed, designed, and optimized for NASA's high-voltage solar array missions.

  15. Cask system design guidance for robotic handling

    International Nuclear Information System (INIS)

    Remote automated cask handling has the potential to reduce both the occupational exposure and the time required to process a nuclear waste transport cask at a handling facility. The ongoing Advanced Handling Technologies Project (AHTP) at Sandia National Laboratories is described. AHTP was initiated to explore the use of advanced robotic systems to perform cask handling operations at handling facilities for radioactive waste, and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof-of-concept robotic systems developed in AHTP are intended to extrapolate from currently available commercial systems to the systems that will be available by the time that a repository would be open for operation. The project investigates those cask handling operations that would be performed at a nuclear waste repository facility during cask receiving and handling. The ongoing AHTP indicates that design guidance, rather than design specification, is appropriate, since the requirements for robotic handling do not place severe restrictions on cask design but rather focus on attention to detail and design for limited dexterity. The cask system design features that facilitate robotic handling operations are discussed, and results obtained from AHTP design and operation experience are summarized. The application of these design considerations is illustrated by discussion of the robot systems and their operation on cask feature mock-ups used in the AHTP project. 11 refs., 11 figs

  16. Tritium recovery from ceramic breeder blanket

    International Nuclear Information System (INIS)

    It is known that chemical forms of tritium released from ceramic breeders are T2O and T2. Among issues relevant to the tritium chemical form, tritium inventory is one of the major criteria in the selection of breeder material. The primary purpose of this report is to study the dependence of tritium inventory in a blanket with ceramic solid breeder on the tritium chemical form. In this light, tritium inventory in a Li2O blanket has been evaluated as a function of tritium chemical form under the conditions of the Japanese Fusion Experimental Reactor (FER). It was shown that in a blanket with Li2O as a breeder, which has a strong affinity to water vapor, the inventory due to T2O adsorption becomes quite large. In order to reduce the T2O adsorption inventory, conversion of the tritium chemical form through an isotope exchange reaction with hydrogen added to the sweep gas (T2O + 2 H2 → H2O + 2 HT) has been proposed, and its advantages and problems have been examined. Lithium hydroxide formation and mass transfer, which are considered to be inherent in the Li2O-T2O system and to be critical issues for the feasibility of a Li2O blanket, have been also discussed. (author)

  17. Review: BNL graphite blanket design concepts

    International Nuclear Information System (INIS)

    A review of the Brookhaven National Laboratory (BNL) minimum activity graphite blanket designs is made. Three designs are identified and discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a thick graphite screen (typically 30 cm or greater, depending on type as well as application-experimental power reactor or commercial reactor). Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy. This energy is then either radiated to a secondary blanket with coolant tubes, as in types A and B, or is removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the structural material of the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude by the graphite screen, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma, whatever the degree of radiation damage

  18. Advanced Polymer For Multilayer Insulating Blankets

    Science.gov (United States)

    Haghighat, R. Ross; Shepp, Allan

    1996-01-01

    Polymer resisting degradation by monatomic oxygen undergoing commercial development under trade name "Aorimide" ("atomic-oxygen-resistant imidazole"). Intended for use in thermal blankets for spacecraft in low orbit, useful on Earth in outdoor applications in which sunlight and ozone degrades other plastics. Also used, for example, to make threads and to make films coated with metals for reflectivity.

  19. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    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.)

  20. Aerogel Blanket Insulation Materials for Cryogenic Applications

    Science.gov (United States)

    Coffman, B. E.; Fesmire, J. E.; White, S.; Gould, G.; Augustynowicz, S.

    2009-01-01

    Aerogel blanket materials for use in thermal insulation systems are now commercially available and implemented by industry. Prototype aerogel blanket materials were presented at the Cryogenic Engineering Conference in 1997 and by 2004 had progressed to full commercial production by Aspen Aerogels. Today, this new technology material is providing superior energy efficiencies and enabling new design approaches for more cost effective cryogenic systems. Aerogel processing technology and methods are continuing to improve, offering a tailor-able array of product formulations for many different thermal and environmental requirements. Many different varieties and combinations of aerogel blankets have been characterized using insulation test cryostats at the Cryogenics Test Laboratory of NASA Kennedy Space Center. Detailed thermal conductivity data for a select group of materials are presented for engineering use. Heat transfer evaluations for the entire vacuum pressure range, including ambient conditions, are given. Examples of current cryogenic applications of aerogel blanket insulation are also given. KEYWORDS: Cryogenic tanks, thermal insulation, composite materials, aerogel, thermal conductivity, liquid nitrogen boil-off

  1. A blanket design, apparatus, and fabrication techniques for the mass production of multilayer insulation blankets for the Superconducting Super Collider

    Energy Technology Data Exchange (ETDEWEB)

    Gonczy, J.D.; Boroski, W.N.; Niemann, R.C.; Otavka, J.G.; Ruschman, M.K.; Schoo, C.J.

    1989-09-01

    The multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) consists of full cryostat length assemblies of aluminized polyester film fabricated in the form of blankets and installed as blankets to the 4.5K cold mass and the 20K and 80K thermal radiation shields. Approximately 40,000 MLI blankets will be required in the 10,000 cryogenic devices comprising the SSC accelerator. Each blanket is nearly 17 meters long and 1.8 meters wide. This paper reports the blanket design, an apparatus, and the fabrication method used to mass produce pre-fabricated MLI blankets. Incorporated in the blanket design are techniques which automate quality control during installation of the MLI blankets in the SSC cryostat. The apparatus and blanket fabrication method insure consistency in the mass produced blankets by providing positive control of the dimensional parameters which contribute to the thermal performance of the MLI blanket. By virtue of the fabrication process, the MLI blankets have inherent features of dimensional stability three-dimensional uniformity, controlled layer density, layer-to-layer registration, interlayer cleanliness, and interlayer material to accommodate thermal contraction differences. 11 refs., 6 figs., 1 tab.

  2. NET remote workstation

    International Nuclear Information System (INIS)

    The goal of this NET study was to define the functionality of a remote handling workstation and its hardware and software architecture. The remote handling workstation has to fulfill two basic functions: (1) to provide the man-machine interface (MMI), that means the interface to the control system of the maintenance equipment and to the working environment (telepresence) and (2) to provide high level (task level) supporting functions (software tools) during the maintenance work and in the preparation phase. Concerning the man-machine interface, an important module of the remote handling workstation besides the standard components of man-machine interfacing is a module for graphical scene presentation supplementing viewing by TV. The technique of integrated viewing is well known from JET BOOM and TARM control using the GBsim and KISMET software. For integration of equipment dependent MMI functions the remote handling workstation provides a special software module interface. Task level support of the operator is based on (1) spatial (geometric/kinematic) models, (2) remote handling procedure models, and (3) functional models of the equipment. These models and the related simulation modules are used for planning, programming, execution monitoring, and training. The workstation provides an intelligent handbook guiding the operator through planned procedures illustrated by animated graphical sequences. For unplanned situations decision aids are available. A central point of the architectural design was to guarantee a high flexibility with respect to hardware and software. Therefore the remote handling workstation is designed as an open system based on widely accepted standards allowing the stepwise integration of the various modules starting with the basic MMI and the spatial simulation as standard components. (orig./HP)

  3. Contact handle decompositions

    OpenAIRE

    Özbağcı, Burak

    2009-01-01

    We review Giroux’s contact handles and contact handle attachments in dimension three and show that a bypass attachment consists of a pair of contact 1 and 2-handles. As an application we describe explicit contact handle decompositions of infinitely many pairwise non-isotopic overtwisted 3-spheres. We also give an alternative proof of the fact that every compact contact 3-manifold (closed or with convex boundary) admits a contact handle decomposition, which is a result originally due to Giroux.

  4. Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

    International Nuclear Information System (INIS)

    The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li2O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

  5. Design of test blanket system for ITER module testing

    International Nuclear Information System (INIS)

    Test blanket systems to be installed in ITER for developing demo blankets have been investigated. One of the main engineering goals of ITER is to test tritium breeding blankets relevant to a power reactor. The test foreseen on modules include the demonstration of a breeding capability that would lead to tritium self-sufficiency in a reactor and extraction of a high grade heat suitable for electricity generation. To accomplish these goals, several ITER equatorial ports are available to test the test blanket systems, both in the basic performance phase (BPP) and the enhanced performance phase (EPP). Test blanket systems for water-cooled and helium-cooled type DEMO blankets with ceramic breeders, developed in Japan have been designed. The design activities include the neutronics, thermal and hydraulic analyses, and mechanical configuration considering port sharing, cooling systems and tritium recovery systems, and test blanket system compatible with the current ITER design has been developed. (author)

  6. Water-cooled blanket concepts for the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    The primary goal of the Blanket Comparison and Selection Study (BCSS) was to select a limited number of blanket concepts for fusion power reactors, to serve as the focus for the U.S. Department of Energy blanket research and development program. The concepts considered most seriously by the BCSS can be grouped for discussion purposes by coolant: liquid metals and alloys, pressurized water, helium, and nitrate salts. Concepts using pressurized water as the coolant are discussed. Water-cooled concepts using liquid breeders-lithium and 17Li-83Pb (LiPb)-have severe fundamental safety problems. The use of lithium and water in the blanket was considered unacceptable. Initial results of tests at Hanford Engineering Development Laboratory using steam injected into molten LiPb indicate that use of LiPb and water together in a blanket is a very serious concern from the safety standpoint. Key issues for water-cooled blankets with solid tritium breeders (Li2O, or a ternary oxide such as LiAlO2) were identified and examined: reliability against leaks, control of tritium permeation into the coolant, retention of breeder physical integrity, breeder temperature predictability, determination of allowable temperature limits for breeders, and 6Li burnup effects (for LiAlO2). The BCSS's final rankings and associated rationale for all water-cooled concepts are examined. Key issues and factors for tokamak and tandem mirror reactor versions of water-cooled solid breeder concepts are discussed. The reference design for the top-ranked concept-LiAlO2 breeder, ferritic steel structure, and beryllium neutron multiplier-is presented. Finally, some general conclusions for water-cooled blanket concepts are drawn based on the study's results

  7. Research on the market availability of mining machines with specific features (radiation protection, remote handling equipment, electric power drive) for the radioactive waste retrieval; Recherche zur Marktverfuegbarkeit von Bergbaumaschinen mit speziellen Eigenschaften (Strahlenschutz, Fernbedienung, Elektroantrieb) zur Rueckholung der radioaktiven Abfaelle aus der Asse (AP-A14)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-12

    The report has been prepared in the frame of the required fact assessment process concerning the state of the waste packages in the emplacement cabins of the Schachtanlage Asse. The search on the market availability of mining machines with specific features (radiation protection, remote handling equipment, electric power drive) for the radioactive waste retrieval covered the following issues: gangue exploitation and room-and-pillar work, underground hauling and transport, railless special vehicles for underground transportation, continuous conveyors and components, feeding and removal equipment, control techniques communication and navigation. The report includes a summary of results and open questions.

  8. Helping Kids Handle Worry

    Science.gov (United States)

    ... Delight: Melon Smoothie Pregnant? Your Baby's Growth Helping Kids Handle Worry KidsHealth > For Parents > Helping Kids Handle ... master life's challenges, big and small. What Do Kids Worry About? What kids worry about is often ...

  9. Water-cooled lithium-lead blanket

    International Nuclear Information System (INIS)

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The present study examines whether the water-cooled lithium-lead blanket designed for NET can be directly extrapolated to a demonstration (DEMO) reactor. A fundamental requirement of the exercise is that the DEMO design should have a tritium breeding ratio which is higher than that in NET. The water-cooled lithium-lead blanket is discussed with respect to: neutronics design, design parameter survey and thermohydraulics, and engineering design. Results are reported of three-dimensional calculations using the Monte Carlo code MORSE-H to investigate possible neutron leakage between the poloidally disposed breeder tubes, and to determine the global tritium breeding ratio for the final double null machine design. (U.K.)

  10. Novel method for sludge blanket measurements.

    Science.gov (United States)

    Schewerda, J; Förster, G; Heinrichmeier, J

    2014-01-01

    The most widely used methods for sludge blanket measurements are based on acoustic or optic principles. In operation, both methods are expensive and often maintenance-intensive. Therefore a novel, reliable and simple method for sludge blanket measurement is proposed. It is based on the differential pressure measurement in the sludge zone compared with the differential pressure in the clear water zone, so that it is possible to measure the upper and the lower sludge level in a tank. Full-scale tests of this method were done in the secondary clarifier at the waste water treatment plant in Hecklingen, Germany. The result shows a good approximation of the manually measured sludge level. PMID:24569276

  11. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

    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

  12. Stellar model atmospheres with magnetic line blanketing

    CERN Document Server

    Kochukhov, O; Shulyak, D

    2004-01-01

    Model atmospheres of A and B stars are computed taking into account magnetic line blanketing. These calculations are based on the new stellar model atmosphere code LLModels which implements direct treatment of the opacities due to the bound-bound transitions and ensures an accurate and detailed description of the line absorption. The anomalous Zeeman effect was calculated for the field strengths between 1 and 40 kG and a field vector perpendicular to the line of sight. The model structure, high-resolution energy distribution, photometric colors, metallic line spectra and the hydrogen Balmer line profiles are computed for magnetic stars with different metallicities and are discussed with respect to those of non-magnetic reference models. The magnetically enhanced line blanketing changes the atmospheric structure and leads to a redistribution of energy in the stellar spectrum. The most noticeable feature in the optical region is the appearance of the 5200 A depression. However, this effect is prominent only in ...

  13. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author)

  14. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    The economic impact of employing the highly ranked blankets in the Blanket Comparison and Selection Study (BCSS) was evaluated in the context of both a tokamak and a tandem mirror power reactor (TMR). The economic evaluation criterion was determined to be the cost of electricity. The influencing factors that were considered are the direct cost of the blankets and related systems; the annual cost of blanket replacement; and the performance of the blanket, heat transfer, and energy conversion systems. The technical and cost bases for comparison were those of the STARFIRE and Mirror Advanced Reactor Study conceptual design power plants. The economic evaluation results indicated that the nitrate-salt-cooled blanket concept is an economically attractive concept for either reactor type. The water-cooled, solid breeder blanket is attractive for the tokamak and somewhat less attractive for the TMR. The helium-cooled, liquidlithium breeder blanket is the least economically desirable of higher ranked concepts. The remaining self-cooled liquid-metal and the helium-cooled blanket concepts represent moderately attractive concepts from an economic standpoint. These results are not in concert with those found in the other BCSS evaluation areas (engineering feasibility, safety, and research and development (R and D) requirements). The blankets faring well economically had generally lower cost components, lower pumping power requirements, and good power production capability. On the other hand, helium- and lithium-cooled systems were preferred from the standpoints of safety, engineering feasibility, and R and D requirements

  15. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Balaraman Kumar

    2010-06-01

    Full Text Available This paper presents the application of an analytical tool to quantify material consistency of offset printing blankets. Printing blankets are essentially viscoelastic rubber composites of several laminas. High levels of material consistency are expected from rubber blankets for quality print and for quick recovery from smash encountered during the printing process. The present study aims at determining objectively the consistency of printing blankets at three specific torque levels of tension under two distinct stages; 1. under normal printing conditions and 2. on recovery after smash. The experiment devised exhibits a variation in tone reproduction properties of each blanket signifying the levels of inconsistency also in thickness direction. Correlation technique was employed on ink density variations obtained from the blanket on paper. Both blankets exhibited good consistency over three torque levels under normal printing conditions. However on smash the recovery of blanket and its consistency was a function of manufacturing and torque levels. This study attempts to provide a new metrics for failure analysis of offset printing blankets. It also underscores the need for optimising the torque for blankets from different manufacturers.

  16. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Lalitha Jayaraman

    2010-01-01

    Full Text Available This paper presents the application of an analytical tool to quantify material consistency of offset printing blankets. Printing blankets are essentially viscoelastic rubber composites of several laminas. High levels of material consistency are expected from rubber blankets for quality print and for quick recovery from smash encountered during the printing process. The present study aims at determining objectively the consistency of printing blankets at three specific torque levels of tension under two distinct stages; 1. under normal printing conditions and 2. on recovery after smash. The experiment devised exhibits a variation in tone reproduction properties of each blanket signifying the levels of inconsistency also in thicknessdirection. Correlation technique was employed on ink density variations obtained from the blanket on paper. Both blankets exhibited good consistency over three torque levels under normal printing conditions. However on smash the recovery of blanket and its consistency was a function of manufacturing and torque levels. This study attempts to provide a new metrics for failure analysis of offset printing blankets. It also underscores the need for optimizing the torque for blankets from different manufacturers.

  17. Neutronics Assessment of Molten Salt Breeding Blanket Design Options

    International Nuclear Information System (INIS)

    Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be ∼1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be (∼3 kg) over the blanket lifetime of ∼3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, the vacuum vessel is reweldable, and the magnets are adequately shielded. We conclude that molten salt blankets can be designed for fusion power plants with neutronics requirements such as adequate tritium breeding and shielding being satisfied

  18. INDRA: a program system for calculating the neutronics and photonics characteristics of a fusion reactor blanket

    International Nuclear Information System (INIS)

    INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU)

  19. Fissile fuel breeding in DT fusion reactor blankets

    International Nuclear Information System (INIS)

    Results of neutronic evaluations of fissile fuel breeding in a variety of DT fusion hybrid-reactor blankets are presented. The blankets are of the fast-fission or fission-suppressed rather than fission-enhanced designs, i.e. in the blankets considered emphasis is on fissile fuel rather than power production. For 233U breeding, when Li metal is the coolant for the first wall and the graphite moderator and the tritium breeding constituent of the blanket, the number of atoms of 233U produced per fusion in blankets that could be of practical interest is in the range 0.5 - 0.68, with the lower value applying to water-cooled ThO2 fertile fuel, the upper to gas-cooled Th-metal fuel located next to the reactor first wall. Neutron multipliers like Pb or Be can increase the production to about 0.74. For 239Pu breeding, the production ratio in practical blankets is 0.6 - 1.64, with the best results being for gas, Na- or Li-metal-cooled U-metal fuels located adjacent to the first wall (the U is depleted uranium). Gas-cooled U-Th-metal blankets, optimized for 233U breeding, yield 0.76 atoms of 233U and 0.38 atoms of 239Pu. The blanket energy multiplication factors are in the range 1.6 - 2.5 for Th blankets, 2.5 - 9.0 for U blankets and approximately 5.5 for the U-Th-metal blanket. The tritium breeding ratio in all blankets is 1.075. Blankets with other first wall, coolant and tritium breeding constituents are also considered. The fusion power requirements of hybrids that could supply the fuel needs of thorium-burning CANDU power reactors, and the allowed costs for building the hybrids are indicated

  20. Detection of Breeding Blankets Using Antineutrinos

    Science.gov (United States)

    Cogswell, Bernadette; Huber, Patrick

    2016-03-01

    The Plutonium Management and Disposition Agreement between the United States and Russia makes arrangements for the disposal of 34 metric tons of excess weapon-grade plutonium. Under this agreement Russia plans to dispose of its excess stocks by processing the plutonium into fuel for fast breeder reactors. To meet the disposition requirements this fuel would be burned while the fast reactors are run as burners, i.e., without a natural uranium blanket that can be used to breed plutonium surrounding the core. This talk discusses the potential application of antineutrino monitoring to the verification of the presence or absence of a breeding blanket. It is found that a 36 kg antineutrino detector, exploiting coherent elastic neutrino-nucleus scattering and made of silicon, could determine the presence of a breeding blanket at a liquid sodium cooled fast reactor at the 95% confidence level within 90 days. Such a detector would be a novel non-intrusive verification tool and could present a first application of coherent elastic neutrino-nucleus scattering to a real-world challenge.

  1. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m2 and a particle heat flux of 1 MW/m2. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  2. Fuel handling machine and auxiliary systems for a fuel handling cell

    International Nuclear Information System (INIS)

    This working report is an update for as well as a supplement to an earlier fuel handling machine design (Kukkola and Roennqvist 2006). A focus in the earlier design proposal was primarily on the selection of a mechanical structure and operating principle for the fuel handling machine. This report introduces not only a fuel handling machine design but also auxiliary fuel handling cell equipment and its operation. An objective of the design work was to verify the operating principles of and space allocations for fuel handling cell equipment. The fuel handling machine is a remote controlled apparatus capable of handling intensely radiating fuel assemblies in the fuel handling cell of an encapsulation plant. The fuel handling cell is air tight space radiation-shielded with massive concrete walls. The fuel handling machine is based on a bridge crane capable of traveling in the handling cell along wall tracks. The bridge crane has its carriage provided with a carousel type turntable having mounted thereon both fixed and telescopic masts. The fixed mast has a gripper movable on linear guides for the transfer of fuel assemblies. The telescopic mast has a manipulator arm capable of maneuvering equipment present in the fuel handling cell, as well as conducting necessary maintenance and cleaning operations or rectifying possible fault conditions. The auxiliary fuel handling cell systems consist of several subsystems. The subsystems include a service manipulator, a tool carrier for manipulators, a material hatch, assisting winches, a vacuum cleaner, as well as a hose reel. With the exception of the vacuum cleaner, the devices included in the fuel handling cell's auxiliary system are only used when the actual encapsulation process is not ongoing. The malfunctions of mechanisms or actuators responsible for the motion actions of a fuel handling machine preclude in a worst case scenario the bringing of the fuel handling cell and related systems to a condition appropriate for

  3. Water-cooled, fire boom blanket, test and evaluation for system prototype development

    International Nuclear Information System (INIS)

    Initial development of actively cooled fire booms indicated that water-cooled barriers could withstand direct oil fire for several hours with little damage if cooling water were continuously supplied. Despite these early promising developments, it was realized that to build reliable full-scale system for Navy host salvage booms would require several development tests and lengthy evaluations. In this experiment several types of water-cooled fire blankets were tested at the Oil and Hazardous Materials Simulated Test Tank (OHMSETT). After the burn test the blankets were inspected for damage and additional tests were conducted to determine handling characteristics for deployment, recovery, cleaning and maintenance. Test results showed that water-cooled fire boom blankets can be used on conventional offshore oil containment booms to extend their use for controlling large floating-oil marine fires. Results also demonstrated the importance of using thermoset rubber coated fabrics in the host boom to maintain sufficient reserve seam strength at elevated temperatures. The suitability of passively cooled covers should be investigated to protect equipment and boom from indirect fire exposure. 1 ref., 2 tabs., 8 figs

  4. Development of advanced blanket performance under irradiation and system integration through JUPITER-II project

    International Nuclear Information System (INIS)

    The Japan-USA collaborative program, JUPITER-II, has made significant progress in a research program titled 'The irradiation performance and system integration of advanced blanket' through a six-year plan for 2001-2006. The scientific concept of this program is to study the elemental technology in macroscopic system integration for advanced fusion blankets based on an understanding of the relevant mechanics at the microscopic level. The program has four main research emphases: (1)Flibe molten salt system: Flibe handling, reduction-oxidation control by Be and Flibe tritium chemistry; thermofluid flow simulation experiment and numerical analysis. (2)Vanadium /Li system: MHD ceramics coating of vanadium alloys and compatibility with Li; neutron irradiation experiment in Li capsule and radiation creep. (3)SiC/He system: Fabrication of advanced composites and property evaluation; thermomechanics of SiC system with solid breeding materials; neutron irradiation experiment in He capsule at high temperatures. (4)Blanket system modeling: Design-based integration modeling of Flibe system and V/Li system; multiscale materials system modeling including He effects. This paper describes the perspective of the program including the historical background, the organization and facilities, and the task objectives. Important recent results are reviewed

  5. The ITER Blanket System Design Challenge

    International Nuclear Information System (INIS)

    Full text: The blanket system is one of the most technically challenging components of the ITER machine, having to accommodate high heat fluxes from the plasma, large electromagnetic loads during off-normal events and demanding interfaces with many key components (in particular the vacuum vessel and in-vessel coils) and the plasma. Plasma scenarios impose demanding requirements on the blanket in terms of heat fluxes on various areas of the first wall during different phases of operation (inboard and outboard midplane for start-up/shut-down scenarios and the top region close to the secondary X-point during flat top) as well as large electro-magnetic (EM) loads and transient energy deposition during off-normal plasma events (such as disruptions and vertical displacement events (VDE)). The high heat fluxes resulting in some areas have necessitated the use of “enhanced heat flux” panels capable of accommodating an incident heat flux of up to 5 MW/m2 in steady state. The other regions utilize “normal heat flux” panels, which have been developed and tested for a heat flux of the order of 1 — 2 MW/m2. The FW shaping design requires a compromise between the conflicting requirements for accommodation of steady state and transient loads (energy deposition during off-normal events). A shaped surface increases the heat loads which are due to plasma particles following the field lines compared to a perfectly toroidal surface. The blanket provides a major contribution to the shielding of the vacuum vessel and coils. A challenging criterion is the need to limit the integrated heating in the toroidal field coil (TFC) to ∼ 14 kW. This is particularly severe on the inboard leg where approximately 80% of the total nuclear heat on the TFC is deposited. Several design modifications were considered and analyzed to help achieve this, including increasing the inboard blanket radial thickness and reducing the assembly gaps. This paper summarizes the latest progress in the

  6. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

  7. Fast-Breeder-Blanket Project: FBBF. Final report

    International Nuclear Information System (INIS)

    This report is the final report for DOE contract DE-AC02-76ET37237 with the Purdue Fast Breeder Blanket Project. The Project was initiated to investigate the uncertainties in Fast Breeder Reactor blanket calculations. Absolute measurements of key neutron reaction rates, neutron spectra, and gamma-ray energy depositions were made in simulated FBF blankets in the Fast Breeder Blanket Facility (FBBF), a Cf-252 driven subcritical facility. Calculation of the spectra and integral reaction rates were made using methods, computer codes, and cross section data typical of those currently used in the design of FBR's. Comparisons of calculated to experimental integral neutron reaction rates give good agreement at the inner portions of the blanket by diverge to C/E ratios of about 0.65 at the outer edge of the blanket for reactions sensitive to the neutron density

  8. The frontiers of research on fusion blanket technology

    International Nuclear Information System (INIS)

    Current topics concerning blanket technology are reviewed. In the chemical engineering/chemistry area, the qualitative and quantitative effects of mass transfer steps of tritium is important in the understanding of the behavior of bred tritium in the solid breeder blanket system. Such phenomena as adsorption, isotope exchange reactions, and water formation reaction at the grain surface produce profound effects on the behavior of the bred tritium in the blanket. Regarding the liquid system, the physical or chemical properties of Li, Li17Pb83 and Flibe as liquid blanket materials were compared. Some recent studies were introduced regarding tritium recovery from the liquid blanket materials, impurity removal from salts, ceramic coating of structural materials, and the vapor pressure of mixtures of metals or salts. Thermal hydraulic topics in relation to several candidate power reactor concepts are summarized. Emphasis is laid on the simultaneous removal of heat and tritium from the blanket and some aspects of forming effective power cycles are developed. (author)

  9. Advanced handling technology project and implications for cask design

    International Nuclear Information System (INIS)

    This paper describes the results of the ongoing Advanced Handling Technologies Project (AHTP) at Sandia. AHTP was initiated in 1986 to explore the use of advanced robotic systems to perform cask handling operations at radioactive waste handling facilities and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof of concept systems developed in AHTP are intended to extrapolate from currently available commercial systems to those that would be available by the time than an actual repository would be open for operation. These systems provide test facilities for the investigation of the robotic handling of alternate cask design features. The following sections describe (1) the approach used in AHTP to select operations for proof of concept robotic systems and to identify the cask design implications, (2) the separate proof of concept robotic systems developed in AHTP, and (3) preliminary insights into the impact of cask system design features on the feasibility of robotic performance of cask handling operations. The main conclusions from AHTP to date regarding design for remote handling are: (1) incorporation of cask system design features which facilitate robotic cask handling can be achieved with minimal impact on cask functional features, (2) proper cask design allows robotic cask handling operations from manipulation of cask tie-down mechanisms to radiation surveys to be performed quickly and reliably without direct human intervention, and (3) design for remote handling also facilitates manual handling operations

  10. A review of fusion breeder blanket technology, part 1

    International Nuclear Information System (INIS)

    This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

  11. Analysis of deficiencies in fast reactor blanket physics predictions

    International Nuclear Information System (INIS)

    This analysis addresses a deviation between experimental measurements and fast reactor blanket physics predictions. A review of worldwide results reveals that reaction rates in the blanket are underpredicted with the discrepancy increasing with penetration into the blanket. The analysis of this discrepancy involves two parts: quantifying possible error reductions using the most advanced methods and investigating deficiencies in current methodology. The source of these discrepancies was investigated by application of ''state-of-the-art'' group constant generation and flux prediction methodology to flux calculations for the Purdue University Fast Breeder Blanket Facility (FBBF). Refined group constant generation methods yielded a significant reduction in the blanket deviations; however, only about half of the discrepancy can be accounted for in this manner. Transport theory calculations were used to predict the blanket neutron transmission problem. The surprising result is that transport theory predictions utilizing diffusion theory group constants did not improve the blanket results. Transport theory predictions exhibited blanket underpredictions similar to the diffusion theory results. The residual blanket discrepancies not explained using advanced methods require a refinement of the theory. For this purpose an analysis of deficiencies in current methodology was performed

  12. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this project advanced acoustic blankets for improved low frequency interior noise control in aircraft will be developed and demonstrated. The improved...

  13. Setting up and managing a remote maintenance operation for fusion

    International Nuclear Information System (INIS)

    Trying to set up and manage a remote maintenance operation for a thermonuclear fusion reactor is a complex undertaking. There are many problems and challenges which need addressing. This paper tries to guide the reader through this process by composing a list of generic problems and by analysing possible solutions. The first challenge before setting up a remote maintenance operation for a fusion reactor is the systematic analysis of all the remote handling requirements. Based upon this the remote handling concept, including facility layout and equipment, can be defined. The following aspects have to be considered and incorporated into the remote handling concept: - Remote handling task development - Remote handling task logistics and resource management - Command, control and human-machine interface system - Viewing and camera systems - Virtual Reality and Augmented Reality software - Automatic path planning and collision avoidance - Remote transfer of heavy loads - Maintainability of RH Equipment - Reliability, redundant systems and safety - Rationalisation and modularity in both hardware and software - Recovery from failure modes - Condition monitoring and fault detection/prediction - Ability to deal with unforeseen problems Oxford Technologies Ltd has a proven track record in setting up and running the Remote Handling group at the JET Joint Undertaking in Culham, UK. Based on the unique experience gained at JET, Oxford Technologies Ltd also developed the current design and remote handling concept of the ITER Hot Cell during a study in 2004. Examples of both the JET Remote Handling experience and the ITER Hot Cell design and layout are given throughout this paper. (orig.)

  14. Setting up and managing a remote maintenance operation for fusion

    International Nuclear Information System (INIS)

    Trying to set up and manage a remote maintenance operation for a thermonuclear fusion reactor is a complex undertaking. There are many problems and challenges which need addressing. This paper tries to guide the reader through this process by composing a list of generic problems and by analysing possible solutions. The first challenge before setting up a remote maintenance operation for a fusion reactor is the systematic analysis of all the remote handling requirements. Based upon this the remote handling concept, including facility layout and equipment, can be defined. The following aspects have to be considered and incorporated into the remote handling concept: - Remote handling task development. - Remote handling task logistics and resource management. - Command, control and human-machine interface system. - Viewing and camera systems. - Virtual reality and Augmented Reality software. - Automatic path planning and collision avoidance. - Remote transfer of heavy loads. - Maintainability of RH equipment. - Reliability, redundant systems and safety. - Rationalisation and modularity in both hardware and software. - Recovery from failure modes. - Condition monitoring and fault detection/prediction. - Ability to deal with unforeseen problems. Oxford Technologies Ltd. has a proven track record in setting up and running the Remote Handling group at the JET Joint Undertaking in Culham, UK. Based on the unique experience gained at JET, Oxford Technologies Ltd. also developed the current design and remote handling concept of the ITER Hot Cell during a study in 2004. Examples of both the JET remote handling experience and the ITER Hot Cell design and layout are given throughout this paper

  15. Oxygen plasma damage to blanket and patterned ultralow-κ surfaces

    International Nuclear Information System (INIS)

    Oxygen plasma damage to blanket and patterned ultralow-κ (ULK) dielectric surfaces was investigated by examining the effect of plasma species and dielectric materials. Blanket ULK films and patterned structures were treated by O2 plasma in a remote plasma chamber where the ions and radicals from the plasma source can be separately controlled to study their respective roles in the damage process. The plasma damage was characterized by angle resolved x-ray photoelectron spectroscopy, x-ray reflectivity, and Fourier transform infrared spectroscopy. Studies of the angle dependence of oxygen plasma damage to blanket ULK films indicated that damage by ions was anisotropic while that by radicals was isotropic. Ions were found to play an important role in assisting carbon depletion by oxygen radicals on the blanket film surface. More plasma damage was observed with increasing porosity in ultralow-κ films. Probable reaction paths were proposed by analyzing the reaction by-products. Plasma damage to the sidewall of low-κ trenches was examined by electron energy loss (EELS) analysis. The depletion depth of carbon was found to be related to the penetration of radical species into the porous dielectric and the distribution at the sidewall and trench bottom was affected by the trench pattern geometry, i.e., the aspect ratio, which can be correlated with the electron potential distribution and subsequent trajectory of ions. Vapor silylation was applied for dielectric recovery of trench structure and the result was examined by EELS. The trimethylchlorosilane was found to be effective for recovery of the sidewall carbon loss. The recovery was better for loss induced by radical O2 than by hybrid O2 and the difference was attributed to the surface densification by ions limiting the mass transport of vapor chemicals.

  16. Development of advanced blanket materials for solid breeder blanket of fusion reactor

    International Nuclear Information System (INIS)

    The design of advanced solid breeding blanket in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high fluence, and the development of such as advanced blanket materials has been carried out by the cooperation activities among JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by wet process is a reference material as a tritium breeder, but the stability on high temperature has to be improved for application to DEMO blanket. As one of such the improved materials, TiO2-doped Li2TiO3 pebbles were successfully fabricated and TiO2-doped Li2TiO3 has been studied. For the advanced neutron multiplier, the beryllides that have high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that Be12Ti had lower swelling and tritium inventory than that of beryllium metal. The pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. From these activities, the bright prospect was obtained to realize the DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides. (author)

  17. Systematic methodology for estimating direct capital costs for blanket tritium processing systems

    International Nuclear Information System (INIS)

    This paper describes the methodology developed for estimating the relative capital costs of blanket processing systems. The capital costs of the nine blanket concepts selected in the Blanket Comparison and Selection Study are presented and compared

  18. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matricies. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data

  19. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matrices. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data

  20. Fusion reactor blanket: neutronic studies in France

    International Nuclear Information System (INIS)

    The problem of effective tritium regeneration is a crucial issue for the fusion reactor, especially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty analysis. The results of these studies permit us to conclude that it is possible to expect an adequate tritium breeding ratio

  1. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 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.

  2. US DCLL test blanket module design and relevance to DEMO

    International Nuclear Information System (INIS)

    Full text: In the design of Test Blanket Modules (TBMs) for ITER, it is required to provide a design concept that is demonstration power reactor (DEMO) relevant. It should be noted that in the US, DEMO is defined to be a good representation of the first generation fusion power reactor. In order to evaluate the potential of the US TBM design for DEMO, a system evaluation of DEMO design was performed with an improved GA system code, and the physics results were benchmarked to ITER. With the selection of ferritic steel as the structural material, the maximum neutron wall loading is limited to 3 MW/m2. When designed to a 3 GW fusion device the optimum aspect ratio is found to be in the range of 2.5 to 3. Results show that the US dual coolant lead-lithium (DCLL) blanket can satisfy all the DEMO design requirements. On the chamber wall material, for the ITER-TBM design, the design guidance is to apply a 2 mm Be layer onto the plasma facing surface. When extrapolated to the DEMO design, the Be layer will not be suitable due to radiation damage. Similarly, a carbon surface will not be suitable due to high physical and chemical sputtering rates, radiation damage of the material and potential large retention of tritium. Unfortunately, the remaining commonly proposed material, tungsten (W), would suffer radiation damage from alpha particle implantation and, with blistering, W transport to the plasma core could severely limit the core performance. To resolve this potential impasse, different innovative options were evaluated. All high performance tokamak experiments presently use boron or silicon to condition the first wall. To use boron in DEMO, it is found that in-situ boronization will be required in order to maintain a boronized layer on the chamber wall. This boronized layer could also protect the W substrate, while retaining low-Z wall characteristics. Further innovative ideas are being evaluated to handle transient events like ELMs and disruptions. TOPICS: (PPCA) Power

  3. Remote operations in a Fusion Engineering Research Facility (FERF)

    International Nuclear Information System (INIS)

    The proposed Fusion Engineering Research Facility (FERF) has been designed for the test and evaluation of materials that will be exposed to the hostile radiation environment created by fusion reactors. Because the FERF itself must create a very hostile radiation environment, extensive remote handling procedures will be required as part of its routine operations as well as for both scheduled and unscheduled maintenance. This report analyzes the remote-handling implications of a vertical- rather than horizontal-orientation of the FERF magnet, describes the specific remote-handling facilities of the proposed FERF installation and compares the FERF remote-handling system with several other existing and proposed facilities. (U.S.)

  4. ITER blanket manifold system: Integration, assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Alex, E-mail: alex.martin@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Dellopoulos, George [F4E, EU ITER Domestic Agency, Barcelona (Spain); Edwards, Paul; Furmanek, Andreas; Gicquel, Stefan; Macklin, Brian [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Martin, Patrick [RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Merola, Mario; Norman, Mark; Raffray, Rene [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    Highlights: •The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. •-The blanket manifold system has been redesigned to improve leak detection and localization. •-The redesign of the blanket manifold system into a system based on individual pipes has proven to be a major engineering challenge. -- Abstract: The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned.

  5. Development of the Helium Cooled Lithium Lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aiello, G., E-mail: giacomo.aiello@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aubert, J.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The HCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • The new design has been developed with the aim to capitalize on TBM experience in ITER. • A new attachment system for the modules has been proposed. - Abstract: The Helium Cooled Lithium Lead (HCLL) blanket is one of the candidate European blanket concepts selected for the DEMOnstration fusion power plant that should follow ITER. In a fusion power plant, the blanket is one of the key components because of its impact on the plant performance, availability, safety and economics. In 2012, the European Fusion Development Agreement (EFDA) agency issued new specifications for DEMO: this paper describes the work performed to adapt the previous 2007 HCLL-DEMO blanket design to those specifications. A new segmentation has been defined assuming straight surfaces for all blanket modules. Following the Multi Module Segment (MMS) option, all modules are attached to a common back supporting structure which also serves as manifold for Helium and PbLi distribution. A detailed CAD design of the central outboard module has been defined. Thermo-hydraulic and thermo-mechanical analyses on of the First Wall and Breeder Zone have been carried out. For the attachment of the modules to the common backplate, a new solution based on the use of Tie Rods, derived from the design of the corresponding HCLL Test Blanket Module for ITER, has been proposed. This paper also identifies the priorities for further development of the HCLL blanket design.

  6. Heterogeneous structure effect on molten salt blanket neutronics

    Energy Technology Data Exchange (ETDEWEB)

    Grebyonkin, K.F.; Kandiev, Ya.Z.; Malyshkin, G.N.; Orlov, A.I. [Inst. of Technical Pysics, Chelyabinsk (Russian Federation). Dept. of Physics

    1997-09-01

    The report presents the results of the molten salt blanket neutronics calculations performed for researchers of a facility for accelerator-driven transmutation of long-lived radioactive wastes and plutonium conversion. Heterogeneous structure effect on molten salt blanket neutronics was studied through computation. 4 refs., 1 fig., 1 tab.

  7. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    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)

  8. Breeding blanket concepts for fusion and materials requirements

    International Nuclear Information System (INIS)

    This paper summarizes the design and performances of recent breeding blanket concepts and identifies the key material issues associated with them. An assessment of different classes of concepts is carried out by balancing out the potential performance of the concepts with the risk associated with the required material development. Finally, an example strategy for blanket development is discussed

  9. ITER blanket manifold system: Integration, assembly and maintenance

    International Nuclear Information System (INIS)

    Highlights: •The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. •-The blanket manifold system has been redesigned to improve leak detection and localization. •-The redesign of the blanket manifold system into a system based on individual pipes has proven to be a major engineering challenge. -- Abstract: The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned

  10. MIT LMFBR blanket research project. Final summary report

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J.

    1983-08-01

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record.

  11. LMFBR Blanket Physics Project progress report No. 6

    International Nuclear Information System (INIS)

    Progress is summarized in experimental and analytical investigations of the neutronics and photonics of benchmark mockups of LMFBR blankets. During the reporting period work was devoted primarily to a wide range of analytical/numerical investigations, including blanket fuel management/economics studies, evaluation of improved blanket designs, and assessment of state-of-the-art methods for gamma heating calculations. Experimental work included preparations for resumption of MIT Reactor operations, primarily fabrication of improved steel reflector assemblies for blanket mockups, and development of an improved radiophotoluminescent readout device for LiF thermoluminescent detectors. The most significant finding was that the neutronic and economic performance of radial blanket assemblies are essentially independent of core size (rating) for radially-power-flattened cores. Hence the methodology and results of current experiments and calculations should be valid for the large commercial LMFBR's of the future

  12. LMFBR Blanket Physics Project progress report No. 6

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J. (ed.)

    1975-06-30

    Progress is summarized in experimental and analytical investigations of the neutronics and photonics of benchmark mockups of LMFBR blankets. During the reporting period work was devoted primarily to a wide range of analytical/numerical investigations, including blanket fuel management/economics studies, evaluation of improved blanket designs, and assessment of state-of-the-art methods for gamma heating calculations. Experimental work included preparations for resumption of MIT Reactor operations, primarily fabrication of improved steel reflector assemblies for blanket mockups, and development of an improved radiophotoluminescent readout device for LiF thermoluminescent detectors. The most significant finding was that the neutronic and economic performance of radial blanket assemblies are essentially independent of core size (rating) for radially-power-flattened cores. Hence the methodology and results of current experiments and calculations should be valid for the large commercial LMFBR's of the future.

  13. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record

  14. Fast-core thermal-blanket breeder reactor

    International Nuclear Information System (INIS)

    A preliminary assessment of the performance expected from a specific type of FCTB reactor, consisting of a gas-cooled fast system for the core and natural-uranium light-water thermal system for the blanket is reported. Both the core and the blanket use the 238U-Pu fuel cycle. When all the neutrons leaking out of the core reach the blanket, the blanket-to-core power ratio is estimated to be about 1.3. By reducing its water-to-fuel volume ratio below 1.5, the light water blanket can be designed to have a higher ksub(eff), while maintaining an equilibrium fissile fuel content. Compared with conventional FBRs, having the same power output, the FCTB reactor considered offers the following advantages: a lower fissile fuel content, easier and safer control, no need for Pu separation. (B.G.)

  15. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

  16. Automated system for handling tritiated mixed waste

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory (LLNL) is developing a semi system for handling, characterizing, processing, sorting, and repackaging hazardous wastes containing tritium. The system combines an IBM-developed gantry robot with a special glove box enclosure designed to protect operators and minimize the potential release of tritium to the atmosphere. All hazardous waste handling and processing will be performed remotely, using the robot in a teleoperational mode for one-of-a-kind functions and in an autonomous mode for repetitive operations. Initially, this system will be used in conjunction with a portable gas system designed to capture any gaseous-phase tritium released into the glove box. This paper presents the objectives of this development program, provides background related to LLNL's robotics and waste handling program, describes the major system components, outlines system operation, and discusses current status and plans

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

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) project began a new design phase called the Engineering Design Activity (EDA) which started in July 1992. A variety of blanket designs options were analyzed as a part of the U.S. ITER home team blanket option trade-off study (BOTS) which began in May 1993. The options considered were a self-cooled Li/V blanket, a helium cooled Li/V blanket and a water cooled 316 SS nonbreeding shield option. Detailed activation, dose rate and waste disposal rating calculations have been performed for these different ITER blanket design options based on a fluence of 3.0 MWa/m2 and an average neutron wall loading of 2.0 MW/m2. A continuous operation assumption was utilized in the analysis. The results of this work are presented in this conference

  18. 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.

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

    International Nuclear Information System (INIS)

    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 (Li2O) and lithium zirconate (Li2ZrO3) 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

  20. Manufacture of blanket shield modules for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzetto, P. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: Patrick.Lorenzetto@tech.efda.org; Boireau, B. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Boudot, C. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Bucci, P. [CEA, DTEN/S3ME/LMIC, 17 rue des Martyrs, F-38054 Grenoble (France); Furmanek, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Ioki, K. [ITER IT, Boltzmannstr. 2, D-85748 Garching (Germany); Liimatainen, J. [Metso Powdermet, P.O. Box 306, FIN-33101 Tampere (Finland); Peacock, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Sherlock, P. [NNC Ltd., Booths Hall, Knutsford, Cheshire WA16 8QZ (United Kingdom); Taehtinen, S. [VTT Industrial Systems, P.O. Box 1704, Espoo, FIN-02044 VTT (Finland)

    2005-11-15

    A research and development programme for the ITER blanket shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups (small scale and medium scale) and full-scale prototypes of shield blocks (SB) and first wall (FW) panels. The manufacturing feasibility of FW panels has been demonstrated for two copper alloy candidates. Two designs have been developed for the manufacture of the SB, one for a conventional fabrication route and one for a fabrication route based on the hot isostatic press technology. This paper presents the fabrication routes developed in Europe for the manufacture of the ITER Shield modules.

  1. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Aqueous solution blanket using lithium salts such as LiNO3 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)

  2. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations

  3. Spacecraft thermal blanket cleaning: Vacuum bake of gaseous flow purging

    Science.gov (United States)

    Scialdone, John J.

    1990-01-01

    The mass losses and the outgassing rates per unit area of three thermal blankets consisting of various combinations of Mylar and Kapton, with interposed Dacron nets, were measured with a microbalance using two methods. The blankets at 25 deg C were either outgassed in vacuum for 20 hours, or were purged with a dry nitrogen flow of 3 cu. ft. per hour at 25 deg C for 20 hours. The two methods were compared for their effectiveness in cleaning the blankets for their use in space applications. The measurements were carried out using blanket strips and rolled-up blanket samples fitting the microbalance cylindrical plenum. Also, temperature scanning tests were carried out to indicate the optimum temperature for purging and vacuum cleaning. The data indicate that the purging for 20 hours with the above N2 flow can accomplish the same level of cleaning provided by the vacuum with the blankets at 25 deg C for 20 hours, In both cases, the rate of outgassing after 20 hours is reduced by 3 orders of magnitude, and the weight losses are in the range of 10E-4 gr/sq cm. Equivalent mass loss time constants, regained mass in air as a function of time, and other parameters were obtained for those blankets.

  4. Handling Pyrophoric Reagents

    Energy Technology Data Exchange (ETDEWEB)

    Alnajjar, Mikhail S.; Haynie, Todd O.

    2009-08-14

    Pyrophoric reagents are extremely hazardous. Special handling techniques are required to prevent contact with air and the resulting fire. This document provides several methods for working with pyrophoric reagents outside of an inert atmosphere.

  5. Handling Stability of Tractor Semitrailer Based on Handling Diagram

    OpenAIRE

    Ren Yuan-yuan; Zheng Xue-lian; Li Xian-sheng

    2012-01-01

    Handling instability is a serious threat to driving safety. In order to analyze the handling stability of a tractor semitrailer, a handling diagram can be used. In our research, considering the impact of multiple nonsteering rear axles and nonlinear characteristics of tires on vehicle handling stability, the handling equations are developed for description of stability of tractor semi-trailer. Then we obtain handling diagrams so as to study the influence of driving speed, loaded mass, and fif...

  6. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

    2015-11-04

    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and

  7. Development of fusion blanket technology for the DEMO reactor.

    Science.gov (United States)

    Colling, B R; Monk, S D

    2012-07-01

    The viability of various materials and blanket designs for use in nuclear fusion reactors can be tested using computer simulations and as parts of the test blanket modules within the International Thermonuclear Experimental Reactor (ITER) facility. The work presented here focuses on blanket model simulations using the Monte Carlo simulation package MCNPX (Computational Physics Division Los Alamos National Laboratory, 2010) and FISPACT (Forrest, 2007) to evaluate the tritium breeding capability of a number of solid and liquid breeding materials. The liquid/molten salt breeders are found to have the higher tritium breeding ratio (TBR) and are to be considered for further analysis of the self sufficiency timing. PMID:22112596

  8. Molten salt cooling/17Li-83Pb breeding blanket concept

    International Nuclear Information System (INIS)

    A description of a fusion breeding blanket concept using draw salt coolant and static 17Li-83Pb is presented. 17Li-83Pb has high breeding capability and low tritium solubility. Draw salt operates at low pressure and is inert to water. Corrosion, MHD, and tritium containment problems associated with the MARS design are alleviated because of the use of a static LiPb blanket. Blanket tritium recovery is by permeation toward the plasma. A direct contact steam generator is proposed to eliminate some generic problems associated with a tube shell steam generator

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

    International Nuclear Information System (INIS)

    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

  10. Hybrid reactor blankets for constant energy multiplication and flat power distribution

    International Nuclear Information System (INIS)

    Two blanket design difficulties are usually attributed to the blanket neutronic properties: high peak-to-average power density ratio distribution and the variation of the energy multiplication with burnup. This work shows that blankets can be designed to have a constant energy multiplication and a flat power distribution. These features are illustrated for light water hydrid reactor blankets

  11. Basic principles of lead and lead-bismuth eutectic application in blanket of fusion reactors

    International Nuclear Information System (INIS)

    High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease magnetohydrodynamic resistance authors propose to form insulating coatings on internal surface of blanket ducts at any moment of fusion reactor exploitation. It may be achieved easily if lead or lead-bismuth eutectic is used and technology of oxidative potential handling is applied. A number of experiments carried out in NNSTU show the availability of the proposed technology. It bases on formation of the insulating coatings that consist of the oxides of components of the structural materials and of the coolant components. In-situ value of the insulating coatings characteristics ρδ is ∼ 10-5 Ohm·m2 for steels and 5,0x10-6 - 5,0x10-5 Ohm·m2 for vanadium alloys. Thermal cycling is possible during exploitation of a blanket. The experimental research of the insulating coatings properties during thermal cycling have shown that the coatings formed into the lead and lead-bismuth coolants save there insulating properties. Experience of many years is an undoubted advantage of the lead-bismuth coolant and less of the lead coolant in comparison with lithium. Russian Federation possesses of experience of exploitation of the research and industrial facilities, of experience of creation of the pumps, steamgenerators and equipment with heavy liquid metal coolants. The unique experience of designing, assembling and exploitation of the fission reactors with lead-bismuth coolant is also available. The problem of technology of lead and lead-bismuth coolants for power high temperature radioactive facilities has been solved. Accidents, emergency situations such as leakage of steamgenerators or depressurization of gas system in facilities with lead and lead-bismuth coolants have been explored and suppressed. (author)

  12. 18 CFR 284.402 - Blanket marketing certificates.

    Science.gov (United States)

    2010-04-01

    ... effective for an affiliated marketer with respect to transactions involving affiliated pipelines when an affiliated pipeline receives its blanket certificate pursuant to § 284.284. (2) Should a marketer...

  13. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the proposed Phase II research effort is to develop heterogeneous (HG) blankets for improved sound reduction in aircraft structures. Phase I...

  14. Safety and personnel access aspects of low activation fusion blankets

    International Nuclear Information System (INIS)

    The use of silicon carbide and carbon materials for structural applications in fusion reactor first wall and blanket regions has been proposed and a continuing effort spent on the development of the ceramics technology. The advantages identified are an extremely low induced radioactivity inventory, a high temperature operating capability, abundant raw material resource availability, and minimized plasma impurity effects. One of the unique features of the applications of these materials to fusion reactor blanket designs is that no alloying element is needed in order to assure the specified mechanical properties such as occurs in metal alloys. The major source of long term radioactivity in these materials is impurities. The impurity elements and their concentrations carried over to the blanket structure during fabrication can be minimized by proper fabrication procedures and techniques. The safety and personnel access aspects of such fusion blankets in conjunction with the impurity element concentration are the main subjects of this paper

  15. Lightweight IMM Multi-Junction Photovoltaic Flexible Blanket Assembly Project

    Data.gov (United States)

    National Aeronautics and Space Administration — DSS's recently completed successful NASA SBIR Phase 1 program has established a TRL 3/4 classification for an innovative IMM PV Integrated Modular Blanket Assembly...

  16. Impact of prescribed burning on blanket peat hydrology

    OpenAIRE

    Holden, J; Palmer, SM; Johnston, K; Wearing, C.; Irvine, B; Brown, LE

    2015-01-01

    Fire is known to impact soil properties and hydrological flowpaths. However, the impact of prescribed vegetation burning on blanket peatland hydrology is poorly understood. We studied ten blanket peat headwater catchments. Five were subject to prescribed burning, while five were unburnt controls. Within the burnt catchments we studied plots where the last burn occurred ∼2 (B2), 4 (B4), 7 (B7) or greater than 10 years (B10+) prior to the start of measurements. These were compared with plots at...

  17. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

    A review is given of the features and potentialities of cold-blanket systems, with respect to plasma equilibrium, stability, and reactor technology. The treatment is concentrated on quasi-steady magnetized plasmas confined at moderately high beta values. The cold-blanket concept has specific potentialities as a fusion reactor, e.g. in connection with the desired densities and dimensions of full-scale systems, refuelling, as well as ash and impurity removal, and stability. (author)

  18. Blankets for tritium catalyzed deuterium (TCD) fusion reactors

    International Nuclear Information System (INIS)

    The TCD fusion fuel cycle - where the 3He from the D(D,n)3He reaction is transmuted, by neutron capture in the blanket, into tritium which is fed back to the plasma - was recently recognized as being potentially more promising than the Catalyzed Deuterium (Cat-D) fuel cycle for tokamak power reactors. It is the purpose of the present work to assess the feasibility of, and to identify promising directions for designing blankets for TCD fusion reactors

  19. Recent developments in fusion first wall, blanket, and shield technology

    International Nuclear Information System (INIS)

    This brief overview of first wall, blanket and shield technology reviews the changes and trends in important design issues in first wall, blanket and shield design and related technology from the 1970's to the 1980's. The emphasis is on base technology rather than either systems engineering or materials development. The review is limited to the two primary confinement systems, tokamaks and mirrors, and production of electricity as the primary goal for development

  20. FIRST STEP blanket structure and fuel assembly design

    International Nuclear Information System (INIS)

    FIRST STEP (Fusion, Inertial, Reduced Requirement Systems Test for Special Nuclear Material, Tritium, and Energy Production) is an Inertial Confinement Fusion (ICF) plant designed to produce tritium, SNM, and energy using near-term technology. It is an integrated facility that will serve as a test bed for fusion power plant technology. The design of the blanket structure and blanket fuel assembly for wetted-wall FIRST STEP reactors is presented here

  1. Remote connector development study

    International Nuclear Information System (INIS)

    Plutonium-uranium extraction (PUREX) connectors, the most common connectors used at the Hanford site, offer a certain level of flexibility in pipe routing, process system configuration, and remote equipment/instrument replacement. However, these desirable features have inherent shortcomings like leakage, high pressure drop through the right angle bends, and a limited range of available pipe diameters that can be connect by them. Costs for construction, maintenance, and operation of PUREX connectors seem to be very high. The PUREX connector designs include a 90 degree bend in each connector. This increases the pressure drop and erosion effects. Thus, each jumper requires at least two 90 degree bends. PUREX connectors have not been practically used beyond 100 (4 in.) inner diameter. This study represents the results of a survey on the use of remote pipe-connection systems in US and foreign plants. This study also describes the interdependence between connectors, remote handling equipment, and the necessary skills of the operators

  2. Remote connector development study

    Energy Technology Data Exchange (ETDEWEB)

    Parazin, R.J.

    1995-05-01

    Plutonium-uranium extraction (PUREX) connectors, the most common connectors used at the Hanford site, offer a certain level of flexibility in pipe routing, process system configuration, and remote equipment/instrument replacement. However, these desirable features have inherent shortcomings like leakage, high pressure drop through the right angle bends, and a limited range of available pipe diameters that can be connect by them. Costs for construction, maintenance, and operation of PUREX connectors seem to be very high. The PUREX connector designs include a 90{degree} bend in each connector. This increases the pressure drop and erosion effects. Thus, each jumper requires at least two 90{degree} bends. PUREX connectors have not been practically used beyond 100 (4 in.) inner diameter. This study represents the results of a survey on the use of remote pipe-connection systems in US and foreign plants. This study also describes the interdependence between connectors, remote handling equipment, and the necessary skills of the operators.

  3. Electrical connectors for blanket modules in ITER

    International Nuclear Information System (INIS)

    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

  4. TRANSPORT/HANDLING REQUESTS

    CERN Multimedia

    Groupe ST/HM

    2002-01-01

    A new EDH document entitled 'Transport/Handling Request' will be in operation as of Monday, 11th February 2002, when the corresponding icon will be accessible from the EDH desktop, together with the application instructions. This EDH form will replace the paper-format transport/handling request form for all activities involving the transport of equipment and materials. However, the paper form will still be used for all vehicle-hire requests. The introduction of the EDH transport/handling request form is accompanied by the establishment of the following time limits for the various services concerned: 24 hours for the removal of office items, 48 hours for the transport of heavy items (of up to 6 metric tons and of standard road width), 5 working days for a crane operation, extra-heavy transport operation or complete removal, 5 working days for all transport operations relating to LHC installation. ST/HM Group, Logistics Section Tel: 72672 - 72202

  5. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  6. Studies on steps affecting tritium residence time in solid blanket

    International Nuclear Information System (INIS)

    For the self sustaining of CTR fuel cycle, the effective tritium recovery from blankets is essential. This means that not only tritium breeding ratio must be larger than 1.0, but also high recovering speed is required for the short residence time of tritium in blankets. Short residence time means that the tritium inventory in blankets is small. In this paper, the tritium residence time and tritium inventory in a solid blanket are modeled by considering the steps constituting tritium release. Some of these tritium migration processes were experimentally evaluated. The tritium migration steps in a solid blanket using sintered breeding materials consist of diffusion in grains, desorption at grain edges, diffusion and permeation through grain boundaries, desorption at particle edges, diffusion and percolation through interconnected pores to purging stream, and convective mass transfer to stream. Corresponding to these steps, diffusive, soluble, adsorbed and trapped tritium inventories and the tritium in gas phase are conceivable. The code named TTT was made for calculating these tritium inventories and the residence time of tritium. An example of the results of calculation is shown. The blanket is REPUTER-1, which is the conceptual design of a commercial reversed field pinch fusion reactor studied at the University of Tokyo. The experimental studies on the migration steps of tritium are reported. (Kako, I.)

  7. Preliminary Safety Analysis of Korea HCSB Test Blanket Module

    International Nuclear Information System (INIS)

    A Helium Cooled Solid Breeder (HCSB) blanket has been considered as one of the promising blanket for the fusion power demonstration plant. Therefore HCSB Test Blanket Module (TBM) testing in ITER is the most important milestone for the development of the blanket of the DEMO plant. Korea has developed the HCSB TBM with some features such as graphite reflector and simplified flow passage. The objective of this study was to evaluate the thermal and structural integrity of the HCSB TBM under the hypothetical accidental conditions such as cooling pipe break in TBM. The safety analysis was performed under conservative conditions based on the TBM design, which can be assumed by the similarity of the safety analysis of the ITER shielding blanket. Transient analysis model was used to calculate the temperature distribution for Loss of Coolant Accident (LOCA). Simplified analysis conditions were a) simultaneous plasma shutdown and LOCA b) LOCA and then after FW temperature reaches 1150 deg. plasma shutdown. Helium circuit behavior during the different LOCA scenarios was also evaluated. Finally the design modifications based on the analysis result and the related R-and-D of the HCSB blanket design for the application in a DEMO reactor were mentioned. (author)

  8. Thermal hydraulics and mechanics research on fusion blanket system

    International Nuclear Information System (INIS)

    In-vessel components such as Blanket and Divertor in a fusion reactor have a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Experimental Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, there are two kinds of blanket systems will be install. One is a shield blanket, which consists of a first wall (FW) and a block module shielding against neutron flux to a vacuum chamber and a superconducting magnet system. The other blanket system is called as a Test Blanket Module (TBM). TBM is a kind of prototype blanket for a fusion power plant and has functions of breeding of tritium (T) and extraction of energy from fusion plasma. TBM consists of FW and T-breeding / neutron (n)-multiplier zone. A concept of TBM developed by JAEA is water-cooled pebble-bed type, which means that FW and other structures are cooled by pressurized high temperature water and T-breeding / n-multiplier zone consists of multiple layers of pebble bed made of T-breeding and n-multiplier material. This paper describes the status of R and Ds on FW and pebble beds from the view of thermo-hydraulics and mechanics. (author)

  9. Practices of Handling

    DEFF Research Database (Denmark)

    Ræbild, Ulla

    Abstract While few will dispute the idea that fashion designers relate to the notion of the body in their work practice, the actual embodied engagement of the designer, and the role that the personal bodies of the designers play in processes of fashion design, is an underexposed although nascent...... dichotomized idea of design as combined, alternating or parallel processes of thinking and doing. In other words, the notion of handling is not about reflection in or on action, as brought to the fore by Schön (1984), but about reflection as action. Below the methodological macro level of handling, the paper...

  10. Remote operation system for container

    International Nuclear Information System (INIS)

    The present invention provides a remote operation system for conducting operation with operation reaction for the inside of a container filled with water (liquid), such as of inner walls and inner structural materials of a BWR type reactor. Namely, a swimming robot comprises a swimming device swimming in the liquid and an attaching/detaching device for holding/releasing the handling robot. A control device remotely operate the swimming robot and the handling robot by way of a cable. A cable processing device takes up or dispenses the cable. In addition, when the swimming robot grasps the handling robot by the attaching/detaching device, the swimming robot transmits an operation instruction sent from the control device by way of the cable to the handling robot. After the attaching/detaching device of the swimming robot releases the handling robot, the handling robot operates based on the transmitted operation instruction. It is preferable that the handling robot has an adsorptive moving device for moving itself while being adsorbed on the wall surface of the container. (I.S.)

  11. The integrated-blanket-coil concept applied to the poloidal field and blanket systems of a tokamak reactor

    International Nuclear Information System (INIS)

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component. This concept, designated the ''integrated-blanket-coil'' (IBC) concept, is applied to the poloidal field and blanket systems of a tokamak reactor. An examination of resistive power losses in the IBC suggests that these losses can be limited to 10% of the fusion thermal power. By assuming a sandwich construction for the IBC walls, magnetohydrodynamic (MHD)-induced pressure drops and associated pressure stresses are shown to be modest and well below design limits. For the stainless steel reference case examined, the MHD-induced pressure drop was estimated to be about 1/3 MPa and the associated primary membrane stress was estimated to be about 47 MPa. The preliminary analyses indicate that the IBC concept offers promise as a means for making fusion reactors more compact by combining blanket and coil functions in a single component

  12. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

  13. Safety Evaluation of the EVOLVE Blanket Concept

    International Nuclear Information System (INIS)

    This article summarizes the results of the safety evaluation of the Evaporation of Lithium and Vapor Extraction (EVOLVE) W-alloy first wall (FW) and blanket concept. We have analyzed the EVOLVE design response during a confinement bypass accident. A confinement bypass accident was chosen because, based on previous safety studies, this accident can produce environmental releases by breaching the primary radioactive confinement boundary of EVOLVE, which is the EVOLVE vacuum vessel (VV). As a consequence of a bypass accident, air from a room adjoining the reactor enters the plasma chamber by way of a failed VV port. This air reacts with the high temperature metals inside of the VV to release energy in the case of a lithium spill, or to mobilize radioactive material by oxidation, and then transport this material to the environment by natural convection airflow through the failed VV port. We use the MELCOR code to analyze the response of EVOLVE during this accident. Based on these results, the EVOLVE concept can meet the no-evacuation dose goal set by the DOE Fusion Safety Standard if the EVOLVE confinement building ventilation system is closed within two hours of the onset of this accident

  14. Current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phase of the Li-LiH, Li-LiD, and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li3N, Li2O, and Li2C2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g., Li--Al and Li--Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li--M alloys can be estimated from lithium activity data for these alloys

  15. Flow characteristics of the Cascade granular blanket

    International Nuclear Information System (INIS)

    Analysis of a single granule on a rotating cone shows that for the 350 half-angle, double-cone-shaped Cascade chamber, blanket granules will stay against the chamber wall if the rotational speed is 50 rpm or greater. The granules move axially down the wall with a slight (5-mm or less) sinusoidal oscillation in the circumferential direction. Granule chute-flow experiments confirm that two-layered flow can be obtained when the chute is inclined slightly above the granular material angle of repose. The top surface layer is thin and fast moving (supercritical flow). A thick bottom layer moves more slowly (subcritical flow controlled at the exit) with a velocity that increases with distance from the bottom of the chute. This is a desirable velocity profile because in the Cascade chamber about one-third of the fusion energy is deposited in the form of x rays and fusion-fuel-pellet debris in the top surface (inner-radius) layer

  16. Microbiological sampling of spacecraft cabling, antennas, solar panels and thermal blankets

    Science.gov (United States)

    Koukol, R. C.

    1973-01-01

    Sampling procedures and techniques described resulted from various flight project microbiological monitoring programs of unmanned planetary spacecraft. Concurrent with development of these procedures, compatibility evaluations were effected with the cognizant spacecraft subsystem engineers to assure that degradation factors would not be induced during the monitoring program. Of significance were those areas of the spacecraft configuration for which special handling precautions and/or nonstandard sample gathering techniques were evolved. These spacecraft component areas were: cabling, high gain antenna, solar panels, and thermal blankets. The compilation of these techniques provides a historical reference for both the qualification and quantification of sampling parameters as applied to the Mariner Spacecraft of the late 1960's and early 1970's.

  17. Basic principles of lead and lead-bismuth eutectic application in blanket of fusion reactors

    International Nuclear Information System (INIS)

    Full text: One of the main requirements of advanced nuclear-power engineering is inherent safety of power installations. It initiates R and D of heavy liquid metals (lead, lead- bismuth eutectic) application in fission reactors as substitute of sodium. The same requirement makes advisable R and D of the lead and lead-bismuth eutectic application in blanket of fusion reactors as substitute of lithium. High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease MHD-resistance authors propose to form electro-insulating coatings on internal surface of blanket ducts at any moment of fusion reactor exploitation. It may be achieved easily if lead or lead-bismuth eutectic is used and technology of oxidative potential handling is applied. A number of experiments carried out in NNSTU show the availability of the proposed technology. It bases on formation of the insulating coatings that consist of the oxides of components of the structural materials and of the coolant components. In-situ value of the electro-insulating coatings characteristics rd (r - specific resistance of coatings, d - thickness) is ∼ 10-5Ω·m2 for steels and 5, 0x10-6 - 5, 0x10-5Ω·m2 for vanadium alloys. Thermal cycling is possible during exploitation of a blanket. The experimental research of the insulating coatings properties during thermal cycling have shown that the coatings formed into the lead and lead-bismuth coolants save there electro-insulating properties. Experience of many years is an undoubted advantage of the lead-bismuth coolant and less of the lead coolant in comparison with lithium. Russian Federation possesses of experience of exploitation of the research and industrial facilities, of experience of creation of the pumps, steam generators and another equipment with heavy liquid metal coolants. The unique experience of designing, assembling and exploitation of the fission reactors with lead-bismuth coolant is also available. The problem

  18. Colonic potassium handling

    DEFF Research Database (Denmark)

    Sørensen, Mads Vaarby; Matos, Joana E.; Prætorius, Helle;

    2010-01-01

    Homeostatic control of plasma K+ is a necessary physiological function. The daily dietary K+ intake of approximately 100 mmol is excreted predominantly by the distal tubules of the kidney. About 10% of the ingested K+ is excreted via the intestine. K+ handling in both organs is specifically regul...

  19. Safe handling of tritium

    International Nuclear Information System (INIS)

    The main objective of this publication is to provide practical guidance and recommendations on operational radiation protection aspects related to the safe handling of tritium in laboratories, industrial-scale nuclear facilities such as heavy-water reactors, tritium removal plants and fission fuel reprocessing plants, and facilities for manufacturing commercial tritium-containing devices and radiochemicals. The requirements of nuclear fusion reactors are not addressed specifically, since there is as yet no tritium handling experience with them. However, much of the material covered is expected to be relevant to them as well. Annex III briefly addresses problems in the comparatively small-scale use of tritium at universities, medical research centres and similar establishments. However, the main subject of this publication is the handling of larger quantities of tritium. Operational aspects include designing for tritium safety, safe handling practice, the selection of tritium-compatible materials and equipment, exposure assessment, monitoring, contamination control and the design and use of personal protective equipment. This publication does not address the technologies involved in tritium control and cleanup of effluents, tritium removal, or immobilization and disposal of tritium wastes, nor does it address the environmental behaviour of tritium. Refs, figs and tabs

  20. Test of fuel handling machine for Monju in sodium

    International Nuclear Information System (INIS)

    Various types of fuel handling machines were studied, and under-the-plug method of fuel exchange and the fuel handling machine of single turning plug, fixed arm type were selected for the prototype reactor ''Monju'', because the turning plug is relatively small, and the rate of operation, safety, operational ability, maintainability and reliability required for the reactor are satisfied, moreover, the extrapolation to the demonstration reactor was considered. Attention must be paid to the points that the fuel handling machine is very long and invisible from outside, and the smooth operation and endurance in sodium are required for it. The full mock-up testing facility of single turning plug, fixed arm type was installed in 1974, and the full mock-up test has been carried out since 1975 in Oarai. Fuel exchange is carried out at about 6 months intervals in Monju, and about 20 to 30% of core and blanket fuels are exchanged for about one month period. The functions required for the fuel handling machine for Monju, the outline of the testing facility, the schedule of the testing, the items of testing and the results, and the matters to be specially written are described. The full mock-up test in sodium has been carried out for 5 years, and the functions and the endurance have been proved sufficiently. (Kako, I.)

  1. Impact of blanket tritium against the tritium plant of fusion reactor

    International Nuclear Information System (INIS)

    The breeder blanket and the blanket tritium recovery system are tested using test blanket modules during ITER campaign. And then, these are integrated with the tritium plant for the first time at a prototype reactor after ITER. In this work, impact to the tritium plant by integration of the solid breeder blanket was discussed. The method of tritium extraction from the blanket and the choice of the process for breeder blanket interface should be discussed not only from the viewpoint of tritium release but also from the viewpoint of the load of processing. (author)

  2. Remote Research

    CERN Document Server

    Tulathimutte, Tony

    2011-01-01

    Remote studies allow you to recruit subjects quickly, cheaply, and immediately, and give you the opportunity to observe users as they behave naturally in their own environment. In Remote Research, Nate Bolt and Tony Tulathimutte teach you how to design and conduct remote research studies, top to bottom, with little more than a phone and a laptop.

  3. SMS BASED REMOTE CONTROL SYSTEM

    Directory of Open Access Journals (Sweden)

    Reecha Ranjan Singh , Sangeeta Agrawal , Saurabh Kapoor ,S. Sharma

    2011-08-01

    Full Text Available A modern world contains varieties of electronic equipment and systems like: TV, security system, Hi-fi equipment, central heating systems, fire alarm systems, security alarm systems, lighting systems, SET Top Box, AC (Air Conditioner etc., we need to handle, ON/OFF or monitor these electrical devices remotely or to communicate with these but, if you are not at the home or that place and you want to communicate with these device. So the new technology for handled these devices remotely and for communication to required the GSM, mobile technology, SMS (short message service and some hardware resources. SMS based remote control for home appliances is beneficial for the human generation, because mobile is most recently used technology nowadays.

  4. Solid waste handling

    International Nuclear Information System (INIS)

    This study presents estimates of the solid radioactive waste quantities that will be generated in the Separations, Low-Level Waste Vitrification and High-Level Waste Vitrification facilities, collectively called the Tank Waste Remediation System Treatment Complex, over the life of these facilities. This study then considers previous estimates from other 200 Area generators and compares alternative methods of handling (segregation, packaging, assaying, shipping, etc.)

  5. Bulk materials handling review

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-02-15

    The paper provides details of some of the most important coal handling projects and technologies worldwide. It describes development by Aubema Crushing Technology GmbH, Bedeschi, Cimbria Moduflex, DBT, Dynamic Air Conveying Systems, E & F Services, InBulk Technologies, Nord-Sen Metal Industries Ltd., Pebco Inc, Primasonics International Ltd., R.J.S. Silo Clean (International) Ltd., Takraf GmbH, and The ACT Group. 17 photos.

  6. Computerised programming of the Dragon reactor fuel handling operations

    International Nuclear Information System (INIS)

    Two suites of FORTRAN IV computer programs have been written to produce check lists for the operation of the two remote control fuel handling machines of the Dragon Reactor. This document describes the advantages of these programs over the previous manual system of writing check lists, and provides a detailed guide to the programs themselves. (author)

  7. Recent MHD activities for blanket analysis at UPC

    International Nuclear Information System (INIS)

    In the frame of fusion reactor design definition, the detailed analysis of main flow parameters in liquid metal blankets is of utmost interest. Critical aspects are (1) tritium inventories and permeation rates, (2) heat extraction and maximum temperatures for material specifications and (3) MHD pressure drops. The aim of GREENER research group at UPC is to develop a CFD code, based on the OpenFOAM toolbox, able to deal with the main phenomena occurring at blanket channels (MHD coupling, heat transfer and tritium transport) and capable to quantify the above mentioned critical aspects. In parallel, CIMNE research group is developing its own MHD code, mainly focused on algorithm optimisation. The paper summarises the developing tools at each research group and compares their behaviour in a validation step using analytical solutions. In order to expose the applicability of the codes, some simulation results related with the HCLL-ITER/TBM blanket are exposed. Special focus is made on buoyancy flows in U-shaped channels and multi channel effect. Moreover, a preliminary flow analysis related with vertical banana-shape liquid metal channels is discussed, related with a new blanket design that is being considered as a progress of conceptual design refinement of dual-coolant liquid metal blankets (DEMO specifications).

  8. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    In nuclear fusion DEMO reactor, the blanket is required to provide the tritium breeding ratio (TBR) of more than unity by the neutron induced reaction in lithium in the blanket. To provide the TBR of more than unity is critical issue in the development of the blanket. Also in order to develop the blanket with low activation level, the evaluation of the induced activity with high precision is required by taking into account the sequential reactions induced by secondary charged particles. In order to evaluate these issues experimentally, neutronics experiments have been performed by using DT neutrons at JAERI/FNS. From the results of TBR experiment by using the mockup relevant to the DEMO blanket with multilayered structure composed of Be, Li2TiO3 and F82H, it was clarified that the TPR can be evaluated within 10 % uncertainty by using the Monte Carlo calculation. From the results of sequential reactions experiment for the test specimens simulating the cooling water pipe, it was found that the effective cross-sections due to the sequential reactions were increased in a form close to an exponential curve in the cooling water pipe with reducing the distance to the water. (author)

  9. Uranium hexafluoride handling

    International Nuclear Information System (INIS)

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF6 from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride

  10. Handling of Solid Residues

    International Nuclear Information System (INIS)

    The topic of solid residues is specifically of great interest and concern for the authorities, institutions and community that identify in them a true threat against the human health and the atmosphere in the related with the aesthetic deterioration of the urban centers and of the natural landscape; in the proliferation of vectorial transmitters of illnesses and the effect on the biodiversity. Inside the wide spectrum of topics that they keep relationship with the environmental protection, the inadequate handling of solid residues and residues dangerous squatter an important line in the definition of political and practical environmentally sustainable. The industrial development and the population's growth have originated a continuous increase in the production of solid residues; of equal it forms, their composition day after day is more heterogeneous. The base for the good handling includes the appropriate intervention of the different stages of an integral administration of residues, which include the separation in the source, the gathering, the handling, the use, treatment, final disposition and the institutional organization of the administration. The topic of the dangerous residues generates more expectation. These residues understand from those of pathogen type that are generated in the establishments of health that of hospital attention, until those of combustible, inflammable type, explosive, radio-active, volatile, corrosive, reagent or toxic, associated to numerous industrial processes, common in our countries in development

  11. Uranium hexafluoride handling. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1991-12-31

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF{sub 6} from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride. Selected papers were processed separately for inclusion in the Energy Science and Technology Database.

  12. Remote Sensing

    CERN Document Server

    Khorram, Siamak; Koch, Frank H; van der Wiele, Cynthia F

    2012-01-01

    Remote Sensing provides information on how remote sensing relates to the natural resources inventory, management, and monitoring, as well as environmental concerns. It explains the role of this new technology in current global challenges. "Remote Sensing" will discuss remotely sensed data application payloads and platforms, along with the methodologies involving image processing techniques as applied to remotely sensed data. This title provides information on image classification techniques and image registration, data integration, and data fusion techniques. How this technology applies to natural resources and environmental concerns will also be discussed.

  13. EU DEMO blanket concepts safety assessment. Final report of Working Group 6a of the Blanket Concept Selection Exercise

    International Nuclear Information System (INIS)

    The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four blanket concepts under development. Two of them use lithium ceramics, the other two concepts employ an eutectic lead-lithium alloy (Pb-17Li) as breeder material. The two most promising concepts were to select in 1995 for further development. In order to prepare the selection, a Blanket Concept Selection Exercise (BCSE) has been inititated by the participating associations under the auspices of the European Commission. This BCSE has been performed in 14 working groups which, in a comparative evaluation of the four blanket concepts, addressed specific fields. The working group safety addressed the safety implications. This report describes the methodology adopted, the safety issues identified, their comparative evaluation for the four concepts, and the results and conclusions of the working group to be entered into the overall evaluation. There, the results from all 14 working groups have been combined to yield a final ranking as a basis for the selection. In summary, the safety assessment showed that the four European blanket concepts can be considered as equivalent in terms of the safety rating adopted, each concept, however, rendering safety concerns of different quality in different areas which are substantiated in this report. (orig.)

  14. Thermomechanics analysis and optimization for high power density blanket

    International Nuclear Information System (INIS)

    Thermomechanics analysis, i.e. steady thermal analysis and steady thermal stress analysis have been carried out for a high power density blanket. The Fusion Experimental Breeder (FEB) is adopted as the reference reactor. The parts for the blanket module in Pro/ENGINEER were created, then turn to Pro/MECHANICA functionality for thermomechanics analysis. During analysis, the distribution of the power density in the blanket was optimized to be more flat, the arched curvature and rounds of the cooling tube panels were optimized to less stiffness, and the boundary condition at the interface of helium cooling tube panel and manifold chamber was optimized, which is reasonable by using advanced welding processes with electron beam or laser beam in a single pass. To the end, a maximum temperature Tm 350 degree C and a maximum shear stress τm 80 MPa for the helium cooling panels have been shown in the calculations. (authors)

  15. Direct Lit Electrolysis In A Metallic Lithium Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Colon-Mercado, H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Babineau, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Elvington, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Garcia-Diaz, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Teprovich, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Vaquer, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-13

    A process that simplifies the extraction of tritium from molten lithium based breeding blankets was developed.  The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fission/fusion reactors is critical in order to maintained low concentrations.  This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Because of the high affinity of tritium for the blanket, extraction is complicated at the required low levels. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering the hydrogen and deuterium thru an electrolysis step at high temperatures. 

  16. Experience in handling abnormal and emergency situations in PHWR fuel handling system

    International Nuclear Information System (INIS)

    On-power Fuel Handling System of PHWR reactor consists of complicated mechanisms operating in multiple media like heavy water, light water and oil. This remote controlled system is the lifeline of PHWR reactor. The complexity of on-power fuel handling system and the need to continuously improve its performance presents challenges at every step. A large number of innovations, modifications and improvements in the system have been made by the stations, design group and R and D units to meet the challenges of higher refueling rate. Innovations in operating/maintenance practices and the methods to safely retrieve from abnormal/emergency situations in shortest possible time had to be specifically devised from the embryonic stage. A lot of efforts were required to be put in by various agencies to develop and formalise the operating procedures for handling various emergency conditions. The implementation of these procedures required the development of special tools/fixtures which had to be tested and tried out in mock-ups before their actual use. The retrieval from emergency situations like handling of damaged bundles in MAPS in early eighties, bundles dropped in shuttle station in NAPS in 1998 and failure of fuel string to move due to damaged bundles at Kaiga in 2003 are some of the most difficult situations handled over the years.This paper focuses on the challenges faced during handling of Safety-related Events in PHWR Fuel Handling System. It also discusses development of procedures and tooling to retrieve from abnormal situations and various innovations and design improvements to avoid the recurrence of the events. (author)

  17. Development of the water cooled lithium lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aiello, G.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The WCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • Preliminary CAD design of the equatorial outboard module of the WCLL blanket has been developed for DEMO. • Finite elements analyses have been carried out in order to assess the module thermal behavior in the straight part of the module. - Abstract: The water cooled lithium lead (WCLL) blanket, based on near-future technology requiring small extrapolation from present-day knowledge both on physical and technological aspect, is one of the breeding blanket concepts considered as possible candidates for the EU DEMOnstration power plant. In 2012, the EFDA agency issued new specifications for DEMO: this paper describes the work performed to adapt the WCLL blanket design to those specifications. Relatively small modules with straight surfaces are attached to a common Back Supporting Structure housing feeding pipes. Each module features reduced activation ferritic-martensitic steel as structural material, liquid Lithium-Lead as breeder, neutron multiplier and carrier. Water at typical Pressurized Water Reactors (PWR) conditions is chosen as coolant. A preliminary design of the equatorial outboard module has been achieved. Finite elements analyses have been carried out in order to assess the module thermal behavior. Two First Wall (FW) concepts have been proposed, one favoring the thermal efficiency, the other favoring the manufacturability. The Breeding Zone has been designed with C-shaped Double-Walled Tubes in order to minimize the Water/Pb-15.7Li interaction likelihood. The priorities for further development of the WCLL blanket concept are identified in the paper.

  18. Nuclear Performance Analyses for HCPB Test Blanket Modules in ITER

    International Nuclear Information System (INIS)

    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 related efforts currently focus on the design optimisation of suitable Test Blanket Modules (TBM) and associated R-and-D activities. Four different HCPB TBM types are considered for addressing issues related to (i) electromagnetic transients (EM), (ii) neutronics and Tritium (NT), (iii) thermo-mechanical properties of the pebble beds (TM), and (iv) the integral performance of the blanket module (Plant Integration, PI). The lay-out of the NT and the PI modules has been entirely revised to represent the latest HCPB breeder blanket concept for fusion power reactors. A HCPB TBM consists of a steel box with an internal stiffening grid and small breeder units. The stiffening grid forms radially running open cells accommodating the breeder units (BU). The BU consists of a back plate with attached breeder canisters providing space for the breeder pebble beds. The space between the canisters and the stiffening plates is filled with Beryllium pebbles for the neutron multiplication. The latest design assumes two vertically arranged breeder containers per BU with a toroidal bed height of 10 and 24 mm, for NT and PI modules, respectively. Li4SiO4 is assumed as breeder material at 6Li enrichment levels between 40 at % (NT) and 90 at % (PI). This work is devoted to the neutronic, shielding and activation analyses performed recently for NT and PI variants of the HCPB TBM in ITER. The analyses are based on three-dimensional neutronic and activation calculations making use of a 20 degree torus sector model of ITER developed for Monte Carlo calculations with the MCNP code. The model includes a proper representation of the horizontal ITER test blanket port, the water cooled support frame with two integrated HCPB blanket test modules, the radiation shield and the port environment. Monte Carlo

  19. Handling during neonatal intensive care.

    OpenAIRE

    Murdoch, D R; Darlow, B A

    1984-01-01

    The handling received by very low birthweight newborns undergoing intensive care in the first few days of life and the effects of this were studied. Infants were handled an average of 4.3 hours (18%) of the total 24 hour observation time and received a mean 234 handling procedures. Parental handling contributed 35% of the total time but was usually benign except in that it could interfere with the infant's rest. Many procedures were associated with undesirable consequences. Endotracheal sucti...

  20. Safe handling of radiopharmaceuticals

    International Nuclear Information System (INIS)

    There has been tremendous growth in the number of industrial and medical institutions using radioisotopes produced in nuclear reactors. This is associated with potential hazards and dangers if appropriate precautions are not taken in handling these. Hence, an effective and meaningful control programme is a must to regulate situations involving potential exposure to radiation such as the production, uses, transport, storage and disposal of radioisotopes. Such regulatory control is generally aimed at protecting the workers and the public from dangers or risks related to ionizing radiation taking into account the net benefit derived. (author)