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Sample records for fuel handling machine

  1. VVER NPPs fuel handling machine control system

    International Nuclear Information System (INIS)

    Mini, G.; Rossi, G.; Barabino, M.; Casalini, M.

    2002-01-01

    In order to increase the safety level of the fuel handling machine on WWER NPPs, Ansaldo Nucleare was asked to design and supply a new Control System. Two Fuel Handling Machine (FHM) Control System units have been already supplied for Temelin NPP and others supply are in process for the Atommash company, which has in charge the supply of FHMs for NPPs located in Russia, Ukraine and China.The computer-based system takes into account all the operational safety interlocks so that it is able to avoid incorrect and dangerous manoeuvres in the case of operator error. Control system design criteria, hardware and software architecture, and quality assurance control, are in accordance with the most recent international requirements and standards, and in particular for electromagnetic disturbance immunity demands and seismic compatibility. The hardware architecture of the control system is based on ABB INFI 90 system. The microprocessor-based ABB INFI 90 system incorporates and improves upon many of the time proven control capabilities of Bailey Network 90, validated over 14,000 installations world-wide.The control system complies all the former designed sensors and devices of the machine and markedly the angular position measurement sensors named 'selsyn' of Russian design. Nevertheless it is fully compatible with all the most recent sensors and devices currently available on the market (for ex. Multiturn absolute encoders).All control logic were developed using standard INFI 90 Engineering Work Station, interconnecting blocks extracted from an extensive SAMA library by using a graphical approach (CAD) and allowing and easier intelligibility, more flexibility and updated and coherent documentation. The data acquisition system and the Man Machine Interface are implemented by ABB in co-operation with Ansaldo. The flexible and powerful software structure of 1090 Work-stations (APMS - Advanced Plant Monitoring System, or Tenore NT) has been successfully used to interface the

  2. WWER NPPs fuel handling machine control system

    International Nuclear Information System (INIS)

    Mini, G.; Rossi, G.; Barabino, M.; Casalini, M.

    2001-01-01

    In order to increase the safety level of the fuel handling machine on WWER NPPs, Ansaldo Nucleare was asked to design and supply a new Control System. Two FHM Control System units have been already supplied for Temelin NPP and others supplies are in process for the Atommash company, which has in charge the supply of FHMs for NPPs located in Russia, Ukraine and China. The Fuel Handling Machine (FHM) Control System is an integrated system capable of a complete management of nuclear fuel assemblies. The computer-based system takes into account all the operational safety interlocks so that it is able to avoid incorrect and dangerous manoeuvres in the case of operator error. Control system design criteria, hardware and software architecture, and quality assurance control, are in accordance with the most recent international requirements and standards, and in particular for electromagnetic disturbance immunity demands and seismic compatibility. The hardware architecture of the control system is based on ABB INFI 90 system. The microprocessor-based ABB INFI 90 system incorporates and improves upon many of the time proven control capabilities of Bailey Network 90, validated over 14,000 installations world-wide. The control system complies all the former designed sensors and devices of the machine and markedly the angular position measurement sensors named 'selsyn' of Russian design. Nevertheless it is fully compatible with all the most recent sensors and devices currently available on the market (for ex. Multiturn absolute encoders). All control logic components were developed using standard INFI 90 Engineering Work Station, interconnecting blocks extracted from an extensive SAMA library by using a graphical approach (CAD) and allowing an easier intelligibility, more flexibility and updated and coherent documentation. The data acquisition system and the Man Machine Interface are implemented by ABB in co-operation with Ansaldo. The flexible and powerful software structure

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

    International Nuclear Information System (INIS)

    Suikki, M.

    2013-10-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Suikki, M. [Optimik Oy, Turku (Finland)

    2013-10-15

    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

  5. Test of fuel handling machine for Monju in sodium

    International Nuclear Information System (INIS)

    Ishii, Yoichiro; Masuda, Yoichi; Kataoka, Hajime

    1980-01-01

    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.)

  6. Review of the Conceptual Design for In-Vessel Fuel Handling Machines in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The main in-vessel fuel handling machines in sodium cooled fast reactor(SFR) are composed of the in-vessel transfer machine(IVTM) and the rotating plug. These machines perform the function to handle fuel assemblies inside the reactor core during the refueling time. The IVTM should be able to access all areas above the reactor core and the fuel transfer port which can discharge the fuel assembly by the rotation of the rotating plug. In the 600 MWe demonstration reactor, the conceptual design of the in-vessel fuel handling machines was carried out. As shown in Fig. 1, the invessel fuel handling machines of the demonstration reactor are the double rotating plug type. With reference to the given core configuration of the demonstration reactor, the arrangement design of the rotating plug was carried out by using the developed simulation program. At present, the conceptual design of SFR prototype reactor which has small capacity of about 100 MWe is being started. Thus, it is necessary the economical efficiency and the reliability of the in-vessel fuel handling machines are reviewed according to the reduction of the power capacity. In this study, the preliminary design concepts of the main invessel fuel handling machines according to the fuel handling type are compared. Also, the design characteristics for the driving mechanism of the IVTM in the demonstration reactor and the recovery concept from the malfunction are reviewed

  7. Nuclear fuel handling apparatus

    International Nuclear Information System (INIS)

    Andrea, C.; Dupen, C.F.G.; Noyes, R.C.

    1977-01-01

    A fuel handling machine for a liquid metal cooled nuclear reactor in which a retractable handling tube and gripper are lowered into the reactor to withdraw a spent fuel assembly into the handling tube. The handling tube containing the fuel assembly immersed in liquid sodium is then withdrawn completely from the reactor into the outer barrel of the handling machine. The machine is then used to transport the spent fuel assembly directly to a remotely located decay tank. The fuel handling machine includes a decay heat removal system which continuously removes heat from the interior of the handling tube and which is capable of operating at its full cooling capacity at all times. The handling tube is supported in the machine from an articulated joint which enables it to readily align itself with the correct position in the core. An emergency sodium supply is carried directly by the machine to provide make up in the event of a loss of sodium from the handling tube during transport to the decay tank. 5 claims, 32 drawing figures

  8. As-Built Verification Plan Spent Nuclear Fuel Canister Storage Building MCO Handling Machine

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    This as-built verification plan outlines the methodology and responsibilities that will be implemented during the as-built field verification activity for the Canister Storage Building (CSB) MCO HANDLING MACHINE (MHM). This as-built verification plan covers THE ELECTRICAL PORTION of the CONSTRUCTION PERFORMED BY POWER CITY UNDER CONTRACT TO MOWAT. The as-built verifications will be performed in accordance Administrative Procedure AP 6-012-00, Spent Nuclear Fuel Project As-Built Verification Plan Development Process, revision I. The results of the verification walkdown will be documented in a verification walkdown completion package, approved by the Design Authority (DA), and maintained in the CSB project files

  9. Remote handling machines

    International Nuclear Information System (INIS)

    Sato, Shinri

    1985-01-01

    In nuclear power facilities, the management of radioactive wastes is made with its technology plus the automatic techniques. Under the radiation field, the maintenance or aid of such systems is important. To cope with this situation, MF-2 system, MF-3 system and a manipulator system as remote handling machines are described. MF-2 system consists of an MF-2 carrier truck, a control unit and a command trailer. It is capable of handling heavy-weight objects. The system is not by hydraulic but by electrical means. MF-3 system consists of a four-crawler truck and a manipulator. The truck is versatile in its posture by means of the four independent crawlers. The manipulator system is bilateral in operation, so that the delicate handling is made possible. (Mori, K.)

  10. Fuel-handling machine tests at the Institute for Nuclear Research - Pitesti. Computer and software research and engineering

    International Nuclear Information System (INIS)

    Doca, Cezar; Predescu, Darie; Maiorescu, Oliviu; Dobrescu, Sorin

    2003-01-01

    This poster introduces the fuel-handling machine SCC-MID. This work is part of a very ambitious project that was accomplished with remarkable investment efforts. Material and human resources was spent to build a test stand for fuel handling machines (CANDU system), closely linked to NPP Cernavoda. A challenging goal was also to develop a computer system (hw/sw) designed and engineered to control the test and calibration process of this fuel-handling machine. The design takes care both of the functionality required to correctly control the F/H machine and of the additional functionality required to assist the testing process. How to test the system itself to validate the implemented solutions, how to safely and consistently maintain the data involved, how to manage such a system, how to gradually integrate the system in the whole stand saving time and work already done and solutions already validated were questions we had to find out right solutions. We choose modular solutions both for hardware and software, based on late technologies which in addition permit to achieve the versatility we needed, namely: VME based hw systems running OS9/68k (Unix like real-time multi-user multitasking OS), ISaGRAF (process control application oriented development and run-time software), Hawk (cross-compiler and IDE software for C/C++ software development intended to run on other Motorola based hw), Suretrack (project management software). The system topology implements open system network concepts that permit communication between different sw/hw platforms (OS9/Motorola and ix86/ms-windows based systems) We spent major resources to model the technological processes and test tools like: - real time simulation of the machine behavior while responding to the human commands or to the state changes of other machine parts as a result of other commands or as mechanical interlinks or technological interlocks and presentation of results revealing time related movements; - database

  11. Testing of FFTF fuel handling equipment

    International Nuclear Information System (INIS)

    Coleman, D.W.; Grazzini, E.D.; Hill, L.F.

    1977-07-01

    The Fast Flux Test Facility has several manual/computer controlled fuel handling machines which are exposed to severe environments during plant operation but still must operate reliably when called upon for reactor refueling. The test programs for two such machines--the Closed Loop Ex-Vessel Machine and the In-Vessel Handling Machine--are described. The discussion centers on those areas where design corrections or equipment repairs substantiated the benefits of a test program prior to plant operation

  12. Fuel handling problems at KANUPP

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, I; Mazhar Hasan, S; Mugtadir, A [Karachi Nuclear Power Plant (KANUPP), Karachi (Pakistan)

    1991-04-01

    KANUPP experienced two abnormal fuel and fuel handling related problems during the year 1990. One of these had arisen due to development of end plate to end plate coupling between the two bundles at the leading end of the fuel string in channel HO2-S. The incident occurred when attempts were being made to fuel this channel. Due to pulling of sticking bundles into the acceptor fuelling machine (north) magazine, which was not designed to accommodate two bundles, a magazine rotary stop occurred. The forward motion of the charge tube was simultaneously discovered to be restricted. The incident led to stalling of fuelling machine locked on to the channel HO2, necessitating a reactor shut down. Removal of the fuelling machine was accomplished sometime later after draining of the channel. The second incident which made the fuelling of channel KO5-N temporarily inexecutable, occurred during attempts to remove its north end shield plug when this channel came up for fuelling. The incident resulted due to breaking of the lugs of the shield plug, making its withdrawal impossible. The Plant however kept operating with suspended fuelling of channel KO5, until it could no longer sustain a further increase in fuel burnup at the maximum rating position. Resolving both these problems necessitated draining of the respective channels, leaving the resident fuel uncovered for the duration of the associated operation. Due to substantial difference in the oxidation temperatures Of UO{sub 2} and Zircaloy and its influence as such on the cooling requirement, it was necessary either to determine explicitly that the respective channels did not contain defective fuel bundles or wait for time long enough to allow the decay heat to reduce to manageable proportions. This had a significant bearing on the Plant down time necessary for the rectification of the problems. This paper describes the two incidents in detail and dwells upon the measures adopted to resolve the related problems. (author)

  13. Fuel handling problems at KANUPP

    International Nuclear Information System (INIS)

    Ahmed, I.; Mazhar Hasan, S.; Mugtadir, A.

    1991-01-01

    KANUPP experienced two abnormal fuel and fuel handling related problems during the year 1990. One of these had arisen due to development of end plate to end plate coupling between the two bundles at the leading end of the fuel string in channel HO2-S. The incident occurred when attempts were being made to fuel this channel. Due to pulling of sticking bundles into the acceptor fuelling machine (north) magazine, which was not designed to accommodate two bundles, a magazine rotary stop occurred. The forward motion of the charge tube was simultaneously discovered to be restricted. The incident led to stalling of fuelling machine locked on to the channel HO2, necessitating a reactor shut down. Removal of the fuelling machine was accomplished sometime later after draining of the channel. The second incident which made the fuelling of channel KO5-N temporarily inexecutable, occurred during attempts to remove its north end shield plug when this channel came up for fuelling. The incident resulted due to breaking of the lugs of the shield plug, making its withdrawal impossible. The Plant however kept operating with suspended fuelling of channel KO5, until it could no longer sustain a further increase in fuel burnup at the maximum rating position. Resolving both these problems necessitated draining of the respective channels, leaving the resident fuel uncovered for the duration of the associated operation. Due to substantial difference in the oxidation temperatures Of UO 2 and Zircaloy and its influence as such on the cooling requirement, it was necessary either to determine explicitly that the respective channels did not contain defective fuel bundles or wait for time long enough to allow the decay heat to reduce to manageable proportions. This had a significant bearing on the Plant down time necessary for the rectification of the problems. This paper describes the two incidents in detail and dwells upon the measures adopted to resolve the related problems. (author)

  14. Sophisticated fuel handling system evolved

    International Nuclear Information System (INIS)

    Ross, D.A.

    1988-01-01

    The control systems at Sellafield fuel handling plant are described. The requirements called for built-in diagnostic features as well as the ability to handle a large sequencing application. Speed was also important; responses better than 50ms were required. The control systems are used to automate operations within each of the three main process caves - two Magnox fuel decanners and an advanced gas-cooled reactor fuel dismantler. The fuel route within the fuel handling plant is illustrated and described. ASPIC (Automated Sequence Package for Industrial Control) which was developed as a controller for the plant processes is described. (U.K.)

  15. Remote filter handling machine for Sizewell B

    International Nuclear Information System (INIS)

    Barker, D.

    1993-01-01

    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)

  16. Fuel transfer machine

    International Nuclear Information System (INIS)

    Bernstein, I.

    1978-01-01

    A nuclear fuel transfer machine for transferring fuel assemblies through the fuel transfer tube of a nuclear power generating plant containment structure is described. A conventional reversible drive cable is attached to the fuel transfer carriage to drive it horizontally through the tube. A shuttle carrying a sheave at each end is arranged in parallel with the carriage to also travel into the tube. The cable cooperating with the sheaves permit driving a relatively short fuel transfer carriage a large distance without manually installing sheaves or drive apparatus in the tunnel. 8 claims, 3 figures

  17. CANDU-9/480-SEU fuel handling system assessment document

    International Nuclear Information System (INIS)

    Hwang, Jeong Ki; Jo, C. H.; Kim, H. M.; Morikawa, D. T.

    1996-11-01

    This report summarize the rationale for the CANDU 9 fuel handling system, and the design choices recommended for components of the system. Some of the design requirements applicable to the CANDU 9 480-SEU fuel handling design choices are described. These requirements imposed by the CANDU 9 project. And the design features for the key components of fuel handling system, such as the fuelling machine, the carriage, the new fuel transfer system and the irradiated fuel transfer system, are described. The carriage seismic load evaluations relevant to the design are contained in the appendices. The majority of the carriage components are acceptable, or will likely be acceptable with some redesign. The concept for the CANDU 9 fuel handling system is based on proven CANDU designs, or on improved CANDU technology. Although some development work must be done, the fuel handling concept is judged to be feasible for the CANDU 9 480-SEU reactor. (author). 2 refs

  18. Recent fuel handling experience in Canada

    International Nuclear Information System (INIS)

    Welch, A.C.

    1991-01-01

    For many years, good operation of the fuel handling system at Ontario Hydro's nuclear stations has been taken for granted with the unavailability of the station arising from fuel handling system-related problems usually contributing less than one percent of the total unavailability of the stations. While the situation at the newer Hydro stations continues generally to be good (with the specific exception of some units at Pickering B) some specific and some general problems have caused significant loss of availability at the older plants (Pickering A and Bruce A). Generally the experience at the 600 MWe units in Canada has also continued to be good with Point Lepreau leading the world in availability. As a result of working to correct identified deficiencies, there were some changes for the better as some items of equipment that were a chronic source of trouble were replaced with improved components. In addition, the fuel handling system has been used three times as a delivery system for large-scale non destructive examination of the pressure tubes, twice at Bruce and once at Pickering and performing these inspections this way has saved many days of reactor downtime. Under COG there are several programs to develop improved versions of some of the main assemblies of the fuelling machine head. This paper will generally cover the events relating to Pickering in more detail but will describe the problems with the Bruce Fuelling Machine Bridges since the 600 MW 1P stations have a bridge drive arrangement that is somewhat similar to Bruce

  19. Fuel Handling Facility Description Document

    International Nuclear Information System (INIS)

    M.A. LaFountain

    2005-01-01

    The purpose of the facility description document (FDD) is to establish the requirements and their bases that drive the design of the Fuel Handling Facility (FHF) to allow the design effort to proceed to license application. This FDD is a living document that will be revised at strategic points as the design matures. It identifies the requirements and describes the facility design as it currently exists, with emphasis on design attributes provided to meet the requirements. This FDD was developed as an engineering tool for design control. Accordingly, the primary audience and users are design engineers. It leads the design process with regard to the flow down of upper tier requirements onto the facility. Knowledge of these requirements is essential to performing the design process. It trails the design with regard to the description of the facility. This description is a reflection of the results of the design process to date

  20. 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

  1. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    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

  2. PND fuel handling decontamination: facilities and techniques

    International Nuclear Information System (INIS)

    Pan, R.Y.

    1996-01-01

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at Ontario Hydro's Pickering Nuclear Division. This paper presents an overview of the set up and techniques used for decontamination in the PND Fuel Handling Maintenance Facility and the effectiveness of each. (author). 1 tab., 9 figs

  3. PND fuel handling decontamination: facilities and techniques

    Energy Technology Data Exchange (ETDEWEB)

    Pan, R Y [Ontario Hydro, Toronto, ON (Canada)

    1997-12-31

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at Ontario Hydro`s Pickering Nuclear Division. This paper presents an overview of the set up and techniques used for decontamination in the PND Fuel Handling Maintenance Facility and the effectiveness of each. (author). 1 tab., 9 figs.

  4. Computerised programming of the Dragon reactor fuel handling operations

    International Nuclear Information System (INIS)

    Butcher, P.

    1976-11-01

    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)

  5. BRET fuel assembly dismantling machine

    International Nuclear Information System (INIS)

    Titzler, P.A.; Bennett, K.L.; Kelley, R.S. Jr.; Stringer, J.L.

    1984-08-01

    An automated remote nuclear fuel assembly milling and dismantling machine has been designed, developed, and demonstrated at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington. The machine can be used to dismantle irradiated breeder fuel assemblies from the Fast Flux Test Facility prior to fuel reprocessing. It can be installed in an existing remotely operated shielded hot cell facility, the Fuels and Materials Examination Facility (FMEF), at the Hanford Site in Richland, Washington

  6. Structural analysis of fuel handling systems

    Energy Technology Data Exchange (ETDEWEB)

    Lee, L S.S. [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1997-12-31

    The purpose of this paper has three aspects: (i) to review `why` and `what` types of structural analysis, testing and report are required for the fuel handling systems according to the codes, or needed for design of a product, (ii) to review the input requirements for analysis and the analysis procedures, and (iii) to improve the communication between the analysis and other elements of the product cycle. The required or needed types of analysis and report may be categorized into three major groups: (i) Certified Stress Reports for design by analysis, (ii) Design Reports not required for certification and registration, but are still required by codes, and (iii) Design Calculations required by codes or needed for design. Input requirements for structural analysis include: design, code classification, loadings, and jurisdictionary boundary. Examples of structural analysis for the fueling machine head and support structure are given. For improving communication between the structural analysis and the other elements of the product cycle, some areas in the specification of design requirements and load rating are discussed. (author). 6 refs., 1 tab., 4 figs.

  7. Structural analysis of fuel handling systems

    International Nuclear Information System (INIS)

    Lee, L.S.S.

    1996-01-01

    The purpose of this paper has three aspects: (i) to review 'why' and 'what' types of structural analysis, testing and report are required for the fuel handling systems according to the codes, or needed for design of a product, (ii) to review the input requirements for analysis and the analysis procedures, and (iii) to improve the communication between the analysis and other elements of the product cycle. The required or needed types of analysis and report may be categorized into three major groups: (i) Certified Stress Reports for design by analysis, (ii) Design Reports not required for certification and registration, but are still required by codes, and (iii) Design Calculations required by codes or needed for design. Input requirements for structural analysis include: design, code classification, loadings, and jurisdictionary boundary. Examples of structural analysis for the fueling machine head and support structure are given. For improving communication between the structural analysis and the other elements of the product cycle, some areas in the specification of design requirements and load rating are discussed. (author). 6 refs., 1 tab., 4 figs

  8. Hoisting appliances and fuel handling equipment at nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-12-31

    The guide is followed by the Finnish Centre for Radiation and Nuclear Safety (STUK) in regulating hoisting and handling equipment Class 3 at nuclear facilities. The guide is applied e.g. to the following equipment: reactor building overhead cranes, hoisting appliances at nuclear fuel storages, fuel handling machines, other hoisting appliances, which because of nuclear safety aspects are classified in Safety Class 3, and load-bearing devices connected with the above equipment, such as replaceable hoisting tools and auxiliary lifting devices. The regulating of hoisting and handling equipment comprises the following stages: handling of preliminary and final safety analysis reports, inspection of the construction plan, supervision of fabrication and construction inspection, and supervision of initial start-up and commissioning inspection. 36 refs. Translation. The original text is published under the same guide number. The guide is valid from 5 January 1987 and will be in force until further notice.

  9. Hoisting appliances and fuel handling equipment at nuclear facilities

    International Nuclear Information System (INIS)

    1987-01-01

    The guide is followed by the Finnish Centre for Radiation and Nuclear Safety (STUK) in regulating hoisting and handling equipment Class 3 at nuclear facilities. The guide is applied e.g. to the following equipment: reactor building overhead cranes, hoisting appliances at nuclear fuel storages, fuel handling machines, other hoisting appliances, which because of nuclear safety aspects are classified in Safety Class 3, and load-bearing devices connected with the above equipment, such as replaceable hoisting tools and auxiliary lifting devices. The regulating of hoisting and handling equipment comprises the following stages: handling of preliminary and final safety analysis reports, inspection of the construction plan, supervision of fabrication and construction inspection, and supervision of initial start-up and commissioning inspection

  10. Human factors issues in fuel handling

    International Nuclear Information System (INIS)

    Beattie, J.D.; Iwasa-Madge, K.M.; Tucker, D.A.

    1994-01-01

    The staff of the Atomic Energy Control Board wish to further their understanding of human factors issues of potential concern associated with fuel handling in CANDU nuclear power stations. This study contributes to that objective by analysing the role of human performance in the overall fuel handling process at Ontario Hydro's Darlington Nuclear Generating Station, and reporting findings in several areas. A number of issues are identified in the areas of design, operating and maintenance practices, and the organizational and management environment

  11. Remote technologies for handling spent fuel

    International Nuclear Information System (INIS)

    Ramakumar, M.S.

    1999-01-01

    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)

  12. Design Package for Fuel Retrieval System Fuel Handling Tool Modification

    International Nuclear Information System (INIS)

    TEDESCHI, D.J.

    2000-01-01

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports

  13. Handling system for nuclear fuel pellet inspection

    International Nuclear Information System (INIS)

    Nyman, D.H.; McLemore, D.R.; Sturges, R.H.

    1978-11-01

    HEDL is developing automated fabrication equipment for fast reactor fuel. A major inspection operation in the process is the gaging of fuel pellets. A key element in the system has been the development of a handling system that reliably moves pellets at the rate of three per second without product damage or excessive equipment wear

  14. Fuel handling grapple for nuclear reactor plants

    International Nuclear Information System (INIS)

    Rousar, D.L.

    1992-01-01

    This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

  15. Powder handling for automated fuel processing

    International Nuclear Information System (INIS)

    Frederickson, J.R.; Eschenbaum, R.C.; Goldmann, L.H.

    1989-01-01

    Installation of the Secure Automated Fabrication (SAF) line has been completed. It is located in the Fuel Cycle Plant (FCP) at the Department of Energy's (DOE) Hanford site near Richland, Washington. The SAF line was designed to fabricate advanced reactor fuel pellets and assemble fuel pins by automated, remote operation. This paper describes powder handling equipment and techniques utilized for automated powder processing and powder conditioning systems in this line. 9 figs

  16. Fuel charging machine

    International Nuclear Information System (INIS)

    Uchikawa, Sadao.

    1978-01-01

    Purpose: To enable continuous fuel discharging and charging steps in a bwr type reactor by effecting positioning only for once by providing a plurality of fuel assembly grippers and their drives co-axially on a rotatable surface. Constitution: A plurality of fuel assembly grippers and their drives are provided co-axially on a rotatable surface. For example, a gripper A, a drive B, a gripper C and a drive D are arranged co-axially in symmetric positions on a disk rotated on rails by wheels and rotational drives. A new fuel in a fuel pool is gripped by the gripper A and transported above the reactor core. Then, the disk is positioned so that the gripper C can grip the spent fuel in the core, and the fuel to be discharged is gripped and raised by the gripper C. Then the disk is rotated by 180 0 and the new fuel in the gripper A is charged into the position from which the old fuel has been discharged and, finally, the discharged fuel is sent to the fuel pool for storage. (Seki, T.)

  17. A Mobile Robot for Emergency Operation of Fuel Exchange Machine

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Yongchil; Lee, Sunguk; Kim, Changhoi; Shin, Hochul; Jung, Seungho; Choi, Changhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    A Pressurized Heavy Water Reactor (PHWR) uses a heavy water as the coolant and moderator because it does not attenuate the neutron inside the reactor, which makes it possible to use natural uranium for nuclear fuels. However, since the uranium ratio is too low within the natural uranium, the reactor should be refueled everyday while the reactor is working. For that purpose, there is a fuel exchange machine. However as the time passes by, the durability and reliability become a problem. While the fuel handling machine exchanges the reactor fuel, it can be stuck to the pressure tube attached in the Calandra. Although this kind of situation is rarely happen, it can make the reactor be shutdown for normalizing the operation. Since the refueling is performed while the reactor is working, the radiation level is extremely high and the machine can be located at a high position up to nine meters from the floor, that is, the human worker can not approach the machine, so the fuel handling machine should be released remotely. To cope with this situation, the fuel handling machine has a manual drive mechanism at the rear side of it as shown in the circled images. If the worker can handle these manual drive mechanisms, the fuel handling machine can be released form the pressure tube. The KAERI had developed a long-reach manipulator system with a telescophic mast mechanism which can be deployed in the basement of the reactor room and manipulate the manual lever of the fuel exchange machine. Since the manipulator is located in the basement, there are several problems for its application such that the plug hole should be removed before the operation and the vibration of the mast mechanism make it difficult to locate the end effecter of the manipulator.

  18. A Mobile Robot for Emergency Operation of Fuel Exchange Machine

    International Nuclear Information System (INIS)

    Seo, Yongchil; Lee, Sunguk; Kim, Changhoi; Shin, Hochul; Jung, Seungho; Choi, Changhwan

    2007-01-01

    A Pressurized Heavy Water Reactor (PHWR) uses a heavy water as the coolant and moderator because it does not attenuate the neutron inside the reactor, which makes it possible to use natural uranium for nuclear fuels. However, since the uranium ratio is too low within the natural uranium, the reactor should be refueled everyday while the reactor is working. For that purpose, there is a fuel exchange machine. However as the time passes by, the durability and reliability become a problem. While the fuel handling machine exchanges the reactor fuel, it can be stuck to the pressure tube attached in the Calandra. Although this kind of situation is rarely happen, it can make the reactor be shutdown for normalizing the operation. Since the refueling is performed while the reactor is working, the radiation level is extremely high and the machine can be located at a high position up to nine meters from the floor, that is, the human worker can not approach the machine, so the fuel handling machine should be released remotely. To cope with this situation, the fuel handling machine has a manual drive mechanism at the rear side of it as shown in the circled images. If the worker can handle these manual drive mechanisms, the fuel handling machine can be released form the pressure tube. The KAERI had developed a long-reach manipulator system with a telescophic mast mechanism which can be deployed in the basement of the reactor room and manipulate the manual lever of the fuel exchange machine. Since the manipulator is located in the basement, there are several problems for its application such that the plug hole should be removed before the operation and the vibration of the mast mechanism make it difficult to locate the end effecter of the manipulator

  19. Human factors issues in fuel handling

    Energy Technology Data Exchange (ETDEWEB)

    Beattie, J D; Iwasa-Madge, K M; Tucker, D A [Humansystems Inc., Milton, ON (Canada)

    1994-12-31

    The staff of the Atomic Energy Control Board wish to further their understanding of human factors issues of potential concern associated with fuel handling in CANDU nuclear power stations. This study contributes to that objective by analysing the role of human performance in the overall fuel handling process at Ontario Hydro`s Darlington Nuclear Generating Station, and reporting findings in several areas. A number of issues are identified in the areas of design, operating and maintenance practices, and the organizational and management environment. 1 fig., 4 tabs., 19 refs.

  20. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    Dazen, J.R.; Denero, J.V.

    1976-01-01

    A nuclear fuel pellet loading machine is described including an inclined rack mounted on a base and having parallel spaced grooves on its upper surface arranged to support fuel rods. A fuel pellet tray is adapted to be placed on a table spaced from the rack, the tray having columns of fuel pellets which are in alignment with the open ends of fuel rods located in the rack grooves. A transition plate is mounted between the fuel rod rack and the fuel pellet tray to receive and guide the pellets into the open ends of the fuel rods. The pellets are pushed into the fuel rods by a number of mechanical fingers mounted on a motor operated block which is moved along the pellet tray length by a drive screw driven by the motor. To facilitate movement of the pellets in the fuel rods the rack is mounted on a number of spaced vibrators which vibrate the fuel rods during fuel pellet insertion. A pellet sensing device movable into an end of each fuel rod indicates to an operator when each rod has been charged with the correct number of pellets

  1. Fuel element load/unload machine for the PEC reactor

    International Nuclear Information System (INIS)

    Clayton, K.F.

    1984-01-01

    GEC Energy Systems Limited are providing two fuel element load/unload machines for use in the Italian fast reactor programme. One will be used in the mechanism test facility (IPM) at Casaccia, to check the salient features of the machine operating in a sodium environment prior to the second machine being installed in the PEC Brasimone Reactor. The machine is used to handle fuel elements, control rods and other reactor components in the sodium-immersed core of the reactor. (U.K.)

  2. Improvements in or relating to gripping means for handling nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Batjukov, V.I.; Vjugov, O.N.; Fadeev, A.I.; Shkhian, T.G.

    1980-01-01

    A gripping means for handling fuel assemblies, the heads of which are internally recessed to receive gripping jaws, forms part of a reactor refuelling machine and is telescopically accommodated within a manipulator tube of the machine. A through hole is provided to allow cooling medium to be passed through the fuel assemblies to remove afterheat when the gripping means is used to transfer assemblies from a reactor core to spent fuel storage sockets. (author)

  3. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Yoon, J. S.; Hong, H. D.; Kim, S. H.

    2004-02-01

    In this research, the remote handling technology is developed for the advanced spent fuel conditioning process which gives a possible solution to deal with the rapidly increasing spent fuels. In detail, a fuel rod slitting device is developed for the decladding of the spent fuel. A series of experiments has been performed to find out the optimal condition of the spent fuel voloxidation which converts the UO 2 pellet into U 3 O 8 powder. The design requirements of the ACP equipment for hot test is established by analysing the modular requirement, radiation hardening and thermal protection of the process equipment, etc. The prototype of the servo manipulator is developed. The manipulator has an excellent performance in terms of the payload to weight ratio that is 30 % higher than that of existing manipulators. To provide reliability and safety of the ACP, the 3 dimensional graphic simulator is developed. Using the simulator the remote handling operation is simulated and as a result, the optimal layout of ACP is obtained. The supervisory control system is designed to control and monitor the several different unit processes. Also the failure monitoring system is developed to detect the possible accidents of the reduction reactor

  4. Numerical simulator of the CANDU fueling machine driving desk

    International Nuclear Information System (INIS)

    Doca, Cezar

    2008-01-01

    As a national and European premiere, in the 2003 - 2005 period, at the Institute for Nuclear Research Pitesti two CANDU fueling machine heads, no.4 and no.5, for the Nuclear Power Plant Cernavoda - Unit 2 were successfully tested. To perform the tests of these machines, a special CANDU fueling machine testing rig was built and was (and is) available for this goal. The design of the CANDU fueling machine test rig from the Institute for Nuclear Research Pitesti is a replica of the similar equipment operating in CANDU 6 type nuclear power plants. High technical level of the CANDU fueling machine tests required the using of an efficient data acquisition and processing Computer Control System. The challenging goal was to build a computer system (hardware and software) designed and engineered to control the test and calibration process of these fuel handling machines. The design takes care both of the functionality required to correctly control the CANDU fueling machine and of the additional functionality required to assist the testing process. Both the fueling machine testing rig and staff had successfully assessed by the AECL representatives during two missions. At same the time, at the Institute for Nuclear Research Pitesti was/is developed a numerical simulator for the CANDU fueling machine operators training. The paper presents the numerical simulator - a special PC program (software) which simulates the graphics and the functions and the operations at the main desk of the computer control system. The simulator permits 'to drive' a CANDU fueling machine in two manners: manual or automatic. The numerical simulator is dedicated to the training of operators who operate the CANDU fueling machine in a nuclear power plant with CANDU reactor. (author)

  5. Man-machine cooperation in remote handling for fusion plants

    International Nuclear Information System (INIS)

    Leinemann, K.

    1984-01-01

    Man-machine cooperation in remote handling for fusion plants comprises cooperation for design of equipment and planning of procedures using a CAD system, and cooperation during operation of the equipment with computer aided telemanipulation systems (CAT). This concept is presently being implemented for support of slave positioning, camera tracking, and camera alignment in the KfK manipulator test facility. The pilot implementation will be used to test various man-machine interface layouts, and to establish a set of basic buildings blocks for future implementations of advanced remote handling control systems. (author)

  6. Experience of safety and performance improvement for fuel handling equipment

    International Nuclear Information System (INIS)

    Gyoon Chang, Sang; Hee Lee, Dae

    2014-01-01

    The purpose of this study is to provide experience of safety and performance improvement of fuel handling equipment for nuclear power plants in Korea. The fuel handling equipment, which is used as an important part of critical processes during the refueling outage, has been improved to enhance safety and to optimize fuel handling procedures. Results of data measured during the fuel reloading are incorporated into design changes. The safety and performance improvement for fuel handling equipment could be achieved by simply modifying the components and improving the interlock system. The experience provided in this study can be useful lessons for further improvement of the fuel handling equipment. (authors)

  7. 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.

  8. Reviewing reactor engineering and fuel handling

    International Nuclear Information System (INIS)

    1991-12-01

    Experience has shown that the better operating nuclear power plants have well defined and effectively administered policies and procedures for governing reactor engineering and fuel handling (RE and FH) activities. This document provides supplementary guidance to OSART experts for evaluating the RE and FH programmes and activities at a nuclear power plant and assessing their effectiveness and adequacy. It is in no way intended to conflict with existing regulations and rules, but rather to exemplify those characteristics and features that are desirable for an effective, well structured RE and FH programme. This supplementary guidance addresses those aspects of RE and FH activities that are required in order to ensure optimum core operation for a nuclear reactor without compromising the limits imposed by the design, safety considerations of the nuclear fuel. In the context of this document, reactor engineering refers to those activities associated with in-core fuel and reactivity management, whereas fuel handling refers to the movement, storage, control and accountability of unirradiated and irradiated fuel. The document comprises five main sections and several appendices. In Section 2 of this guide, the essential aspects of an effective RE and FH programme are discussed. In Section 3, the various types of documents and reference materials needed for the preparatory work and investigation are listed. In Section 4, specific guidelines for investigation of RE and FH programmes are presented. In Section 5, the essential attributes of an excellent RE and FH programme are listed. The supplementary guidance is concluded with a series of appendices exemplifying the various qualities and attributes of a sound, well defined RE and FH programme

  9. Design characteristics of pantograph type in vessel fuel handling system in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

  10. Design characteristics of pantograph type in vessel fuel handling system in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

  11. Safe handling of renewable fuels and fuel mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Wilen, C; Rautalin, A [VTT Energy, Espoo (Finland)

    1997-12-01

    VTT Energy has for several years carried out co-operation with many European research institutes on contractional basis on safety issues of fuels handling. A two-year co-operational project between VTT Energy and these research institutes was started in EU`s JOULE 3 programme in 1996, the total budget of which is 6.9 million FIM. Dust explosion testing method for `difficult` fuels, and for tests at elevated pressures and temperatures, will be developed in the task `Safe handling of renewable fuels and fuel mixtures`. Self- ignition and dust-explosion characteristics will be generated for wood and agro-biomass based biomasses and for the mixtures of them and coal. Inertization requirements will be studied, and the quenching method, combined with partial inertization, will be tested in 1.0 m{sup 3} test equipment. The ignition properties of the fuels under normal and elevated pressures will be characterised with thermobalances. The self-ignition tests with wood and forest residue dusts at 25 bar pressure have been carried out as scheduled. In addition to this, several fuels have undergone thermobalance tests, sieve analyses and microscopic studies for the characterisation of the fuels

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

    International Nuclear Information System (INIS)

    Haist, B.; Hamilton, D.; Sanders, St.

    2006-01-01

    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)

  13. Review on Fuel Loading Process and Performance for Advanced Fuel Handling Equipment

    International Nuclear Information System (INIS)

    Chang, Sang-Gyoon; Lee, Dae-Hee; Kim, Young-Baik; Lee, Deuck-Soo

    2007-01-01

    The fuel loading process and the performance of the advanced fuel handling equipment for OPR 1000 (Optimized Power Plant) are analyzed and evaluated. The fuel handling equipment, which acts critical processes in the refueling outage, has been improved to reduce fuel handling time. The analysis of the fuel loading process can be a useful tool to improve the performance of the fuel handling equipment effectively. Some recommendations for further improvement are provided based on this study

  14. Enhanced wood fuel handling: market and design studies

    Energy Technology Data Exchange (ETDEWEB)

    Landen, R.; Rippengal, R.; Redman, A.N.

    1997-09-01

    This report examines the potential for the manufacture and sale of novel wood fuel handling systems as a means of addressing users' concerns regarding current capital costs and potential high labour costs of non-automated systems. The report considers fuel handling technology that is basically appropriate for wood-fired heating systems of between c.100kW and c.1MW maximum continuous rating. This report details work done by the project collaborators in order to: (1) assess the current status of wood fuel handling technology; (2) evaluate the market appetite for improved wood fuel handling technology; (3) derive capital costs which are acceptable to customers; (4) review design options; and (5) select one or more design options worthy of further development. The current status of wood fuel handling technology is determined, and some basic modelling to give guidance on acceptable capital costs of 100-1000kW wood fuel handling systems is undertaken. (author)

  15. Development of first full scope commercial CANDU-6 fuel handling simulator

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, W., E-mail: BCrawford@atlanticnuclear.ca [Atlantic Nuclear Services Inc., Fredericton, NB (Canada); McInerney, J. M., E-mail: JMcInerney@nbpower.com [Point Lepreau Generating Station, Maces Bay, NB (Canada); Moran, E.S.; Nice, J. W.; Sinclair, D.M.; Somerville, S.; Usalp, E.C.; Usalp, M., E-mail: EMoran@atlanticnuclear.ca, E-mail: JNice@atlanticnuclear.ca, E-mail: DSinclair@atlanticnuclear.ca, E-mail: SSomerville@atlanticnuclear.ca, E-mail: ECUsalp@atlanticnuclear.ca, E-mail: MUsalp@atlanticnuclear.ca [Atlantic Nuclear Services Inc., Fredericton, NB (Canada)

    2015-07-01

    Unique to CANDU reactors is continuous on-power refueling. In the CANDU-6 design, the fuel bundles are contained within 380 pressure tubes. Fuelling machines, one on either side of the reactor face move on a bridge and carriage system to the appointed channel and fuel under computer control. The fuelling machine is an immensely complicated mechanical device. None of the original Canadian full scope simulators incorporated the interaction of the fuel handling system. Traditionally, the final stages of Fuel Handling Operator qualification utilizes on the job training in a production environment carried out in the station main control room. For the purpose of supporting continual improvement in fuel handling training at the Third Qinshan Nuclear Plant Company (TQNPC), Atlantic Nuclear Services in a joint project with New Brunswick Power, developed the first commercial full scope CANDU-6 Fuel Handling simulator, integrated into the existing TQNPC Full Scope Simulator framework. The TQNPC Fuel Handling simulator is capable of supporting all normal on-power and off-power refuelling procedures as well as other abnormal operating conditions, which will allow training to be conducted, based on the plant specific operating procedures. This paper will discuss its development, the importance of this tool and its advantages over past training practices. (author)

  16. Development of first full scope commercial CANDU-6 fuel handling simulator

    International Nuclear Information System (INIS)

    Crawford, W.; McInerney, J. M.; Moran, E.S.; Nice, J. W.; Sinclair, D.M.; Somerville, S.; Usalp, E.C.; Usalp, M.

    2015-01-01

    Unique to CANDU reactors is continuous on-power refueling. In the CANDU-6 design, the fuel bundles are contained within 380 pressure tubes. Fuelling machines, one on either side of the reactor face move on a bridge and carriage system to the appointed channel and fuel under computer control. The fuelling machine is an immensely complicated mechanical device. None of the original Canadian full scope simulators incorporated the interaction of the fuel handling system. Traditionally, the final stages of Fuel Handling Operator qualification utilizes on the job training in a production environment carried out in the station main control room. For the purpose of supporting continual improvement in fuel handling training at the Third Qinshan Nuclear Plant Company (TQNPC), Atlantic Nuclear Services in a joint project with New Brunswick Power, developed the first commercial full scope CANDU-6 Fuel Handling simulator, integrated into the existing TQNPC Full Scope Simulator framework. The TQNPC Fuel Handling simulator is capable of supporting all normal on-power and off-power refuelling procedures as well as other abnormal operating conditions, which will allow training to be conducted, based on the plant specific operating procedures. This paper will discuss its development, the importance of this tool and its advantages over past training practices. (author)

  17. Development of the Simulation Program for the In-Vessel Fuel Handling System of Double Rotating Plug Type

    International Nuclear Information System (INIS)

    Kim, S. H.; Kim, J. B.

    2011-01-01

    In-vessel fuel handling machines are the main equipment of the in-vessel fuel handling system, which can move the core assembly inside the reactor vessel along with the rotating plug during refueling. The in vessel fuel handling machines for an advanced sodium cooled fast reactor(SFR) demonstration plant are composed of a direct lift machine(DM) and a fixed arm machine(FM). These machines should be able to access all areas above the reactor core by means of the rotating combination of double rotating plugs. Thus, in the in vessel fuel handling system of the double rotating plug type, it is necessary to decide the rotating plug size and evaluate the accessibility of in-vessel fuel handling machines in given core configuration. In this study, the simulation program based on LABVIEW which can effectively perform the arrangement design of the in vessel fuel handling system and simulate the rotating plug motion was developed. Fig. 1 shows the flow chart of the simulation program

  18. Full scale tests on remote handled FFTF fuel assembly waste handling and packaging

    International Nuclear Information System (INIS)

    Allen, C.R.; Cash, R.J.; Dawson, S.A.; Strode, J.N.

    1986-01-01

    Handling and packaging of remote handled, high activity solid waste fuel assembly hardware components from spent FFTF reactor fuel assemblies have been evaluated using full scale components. The demonstration was performed using FFTF fuel assembly components and simulated components which were handled remotely using electromechanical manipulators, shielding walls, master slave manipulators, specially designed grapples, and remote TV viewing. The testing and evaluation included handling, packaging for current and conceptual shipping containers, and the effects of volume reduction on packing efficiency and shielding requirements. Effects of waste segregation into transuranic (TRU) and non-transuranic fractions also are discussed

  19. Handling apparatus for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Hornak, L.P.; Desmarchais, W.E.

    1978-01-01

    An apparatus is disclosed for handling radioactive fuel assembly during transfer operations. The radioactive fuel assembly is drawn up into a shielding sleeve which substantially reduces the level of radioactivity immediately surrounding the sleeve thereby permitting direct access by operating personnel. The lifting assembly which draws the fuel assembly up within the shielding sleeve is mounted to and forms an integral part of the handling apparatus. The shielding sleeve accompanies the fuel assembly during all of the transfer operations

  20. Overview of the CANDU fuel handling system for advanced fuel cycles

    International Nuclear Information System (INIS)

    Koivisto, D.J.; Brown, D.R.

    1997-01-01

    Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

  1. Development of machine vision system for PHWR fuel pellet inspection

    Energy Technology Data Exchange (ETDEWEB)

    Kamalesh Kumar, B.; Reddy, K.S.; Lakshminarayana, A.; Sastry, V.S.; Ramana Rao, A.V. [Nuclear Fuel Complex, Hyderabad, Andhra Pradesh (India); Joshi, M.; Deshpande, P.; Navathe, C.P.; Jayaraj, R.N. [Raja Ramanna Centre for Advanced Technology, Indore, Madhya Pradesh (India)

    2008-07-01

    Nuclear Fuel Complex, a constituent of Department of Atomic Energy; India is responsible for manufacturing nuclear fuel in India . Over a million Uranium-di-oxide pellets fabricated per annum need visual inspection . In order to overcome the limitations of human based visual inspection, NFC has undertaken the development of machine vision system. The development involved designing various subsystems viz. mechanical and control subsystem for handling and rotation of fuel pellets, lighting subsystem for illumination, image acquisition system, and image processing system and integration. This paper brings out details of various subsystems and results obtained from the trials conducted. (author)

  2. PND fuel handling decontamination program: specialized techniques and results

    International Nuclear Information System (INIS)

    Pan, R.; Hobbs, K.; Minnis, M.; Graham, K.

    1995-01-01

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at the Pickering Nuclear Station, an eight unit CANDU station located about 30 km east of Toronto. This paper presents an overview of the set up and techniques used for cleaning in the PND Fuel Handling Maintenance Facility, and the results achieved. (author)

  3. Fuel handling and storage systems in nuclear power plants

    International Nuclear Information System (INIS)

    1984-01-01

    The scope of this Guide includes the design of handling and storage facilities for fuel assemblies from the receipt of fuel into the nuclear power plant until the fuel departs from that plant. The unirradiated fuel considered in this Guide is assumed not to exhibit any significant level of radiation so that it can be handled without shielding or cooling. This Guide also gives limited consideration to the handling and storage of certain core components. While the general design and safety principles are discussed in Section 2 of this Guide, more specific design requirements for the handling and storage of fuel are given in detailed sections which follow the general design and safety principles. Further useful information is to be found in the IAEA Technical Reports Series No. 189 ''Storage, Handling and Movement of Fuel and Related Components at Nuclear Power Plants'' and No. 198 ''Guide to the Safe Handling of Radioactive Wastes at Nuclear Power Plants''. However, the scope of the Guide does not include consideration of the following: (1) The various reactor physics questions associated with fuel and absorber loading and unloading into the core; (2) The design aspects of preparation of the reactor for fuel loading (such as the removal of the pressure vessel head for a light water reactor) and restoration after loading; (3) The design of shipping casks; (4) Fuel storage of a long-term nature exceeding the design lifetime of the nuclear power plant; (5) Unirradiated fuel containing plutonium

  4. Arrival condition of spent fuel after storage, handling, and transportation

    International Nuclear Information System (INIS)

    Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

    1982-11-01

    This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables

  5. Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

    International Nuclear Information System (INIS)

    Black, J.E.; King, F.D.

    1979-01-01

    Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

  6. Remote handling technology for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Sakai, Akira; Maekawa, Hiromichi; Ohmura, Yutaka

    1997-01-01

    Design and R and D on nuclear fuel cycle facilities has intended development of remote handling and maintenance technology since 1977. IHI has completed the design and construction of several facilities with remote handling systems for Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan Atomic Energy Research Institute (JAERI), and Japan Nuclear Fuel Ltd. (JNFL). Based on the above experiences, IHI is now undertaking integration of specific technology and remote handling technology for application to new fields such as fusion reactor facilities, decommissioning of nuclear reactors, accelerator testing facilities, and robot simulator-aided remote operation systems in the future. (author)

  7. Fuel handling system of nuclear reactor plants

    International Nuclear Information System (INIS)

    Faulstich, D.L.

    1991-01-01

    This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

  8. Spent fuel cask handling at an operating nuclear power plant

    International Nuclear Information System (INIS)

    Pal, A.C.

    1988-01-01

    The importance of spent fuel handling at operating nuclear power plants cannot be overstated. Because of its highly radioactive nature, however, spent fuel must be handled in thick, lead-lined containers or casks. Thus, all casks for spent fuel handling are heavy loads by the US Nuclear Regulatory Commission's definition, and any load-drop must be evaluated for its potential to damage safety-related equipment. Nuclear Regulatory Guide NUREG-0612 prescribes the regulatory requirements of alternative heavy-load-handling methodologies such as (a) by providing cranes that meet the requirements of NUREG-0554, which shall be called the soft path, or (b) by providing protective devices at all postulated load-drop areas to prevent any damage to safety-related equipment, which shall be called the hard path. The work reported in this paper relates to cask handling at New York Power Authority's James A. FitzPatrick (JAF) plant

  9. Conceptual design report for a Fusion Engineering Device sector-handling machine and movable manipulator system

    International Nuclear Information System (INIS)

    Watts, K.D.; Masson, L.S.; McPherson, R.S.

    1982-10-01

    Design requirements, trade studies, design descriptions, conceptual designs, and cost estimates have been completed for the Fusion Engineering Device sector handling machine, movable manipulator system, subcomponent handling machine, and limiter blade handling machine. This information will be used by the Fusion Engineering Design Center to begin to determine the cost and magnitude of the effort required to perform remote maintenance on the Fusion Engineering Device. The designs presented are by no means optimum, and the costs estimates are rough-order-of-magnitude

  10. Nuclear fuel handling grapple carriage with self-lubricating bearing

    International Nuclear Information System (INIS)

    1977-01-01

    This invention relates to the provision of a fuel handling grapple carriage for a sodium cooled fast breeder reactor with sodium coolant lubricated bearings in which contamination of the bearings is prevented. (UK)

  11. Examples of remote handling of irradiated fuel assemblies in Germany

    International Nuclear Information System (INIS)

    Peehs, M.; Knecht, K.

    1999-01-01

    Examples for the remote handling of irradiated fuel in Germany are presented in the following areas: - fuel assembling pool service activities; - early encapsulation of spent fuel in the pool of a nuclear power plant (NPP) at the end of the wet storage period. All development in remote fuel assembly handling envisages minimization of the radioactive dose applied to the operating staff. In the service area a further key objective for applying advanced methods is to perform the work faster and at a higher quality standard. The early encapsulation is a new technology to provide the final packaging of spent fuel already in the pool of a NPP to ensure reliable handling for all further back end processes. (author)

  12. Safety for fuel assembly handling in the nuclear ship Mutsu

    International Nuclear Information System (INIS)

    Ando, Yoshio

    1978-01-01

    The safety for fuel assembly handling in the nuclear ship Mutsu is deliberated by the committee of general inspection and repair technique examination for Mutsu. The result of deliberation for both cases of removing fuel assemblies and keeping them in the reactor is outlined. The specification of fuel assemblies, and the nuclides and designed radioactivity of fission products of fuel are described. The possibility of shielding repair work and general safety inspection keeping the fuel assemblies in the reactor, the safety consideration when the fuel assemblies are removed at a quay, in a dry dock and on the ocean, the safety of fuel transport in special casks and fuel storage are explained. It is concluded finally that the safety of shielding repair work and general inspection work is secured when the fuel assemblies are kept in the reactor and also when the fuel assemblies are removed from the reactor by cautious working. (Nakai, Y.)

  13. Handling and transfer operations for partially-spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, J K [PUSPATI, Kuala Lumpur (Malaysia)

    1983-12-01

    This project involved the handling and transfer of partially-spent reactor fuel from the Oregon State University TRIGA Reactor in Corvallis, Oregon to Hanford Engineering Development Laboratory in Richland, Washington. The method of handling is dependent upon the burn-up history of the fuel elements. Legal constraints imposed by standing U.S. nuclear regulations determine the selection of transport containers, transportation procedures, physical security arrangements in transit and nuclear material accountability documentation. Results of in-house safety evaluations of the project determine the extent of involvement of pertinent nuclear regulatory authorities. The actual handling activities and actual radiation dose rates are also presented.

  14. Handling and transfer operations for partially-spent nuclear fuel

    International Nuclear Information System (INIS)

    Ibrahim, J.K.

    1983-01-01

    This project involved the handling and transfer of partially-spent reactor fuel from the Oregon State University TRIGA Reactor in Corvallis, Oregon to Hanford Engineering Development Laboratory in Richland, Washington. The method of handling is dependent upon the burn-up history of the fuel elements. Legal constraints imposed by standing U.S. nuclear regulations determine the selection of transport containers, transportation procedures, physical security arrangements in transit and nuclear material accountability documentation. Results of in-house safety evaluations of the project determine the extent of involvement of pertinent nuclear regulatory authorities. The actual handling activities and actual radiation dose rates are also presented (author)

  15. Encapsulation and handling of spent nuclear fuel for final disposal

    International Nuclear Information System (INIS)

    Loennerberg, B.; Larker, H.; Ageskog, L.

    1983-05-01

    The handling and embedding of those metal parts which arrive to the encapsulation station with the fuel is described. For the encapsulation of fuel two alternatives are presented, both with copper canisters but with filling of lead and copper powder respectively. The sealing method in the first case is electron beam welding, in the second case hot isostatic pressing. This has given the headline of the two chapters describing the methods: Welded copper canister and Pressed copper canister. Chapter 1, Welded copper canister, presents the handling of the fuel when it arrives to the encapsulation station, where it is first placed in a buffer pool. From this pool the fuel is transferred to the encapsulation process and thereby separated from fuel boxes and boron glass rod bundles, which are transported together with the fuel. The encapsulation process comprises charging into a copper canister, filling with molten lead, electron beam welding of the lid and final inspection. The transport to and handling in the final repository are described up to the deposition and sealing in the deposition hole. Handling of fuel residues is treated in one of the sections. In chapter 2, Pressed copper canister, only those parts of the handling, which differ from chapter 1 are described. The hot isostatic pressing process is given in the first sections. The handling includes drying, charging into the canister, filling with copper powder, seal lid application and hot isostatic pressing before the final inspection and deposition. In the third chapter, BWR boxes in concrete moulds, the handling of the metal parts, separated from the fuel, are dealt with. After being lifted from the buffer pool they are inserted in a concrete mould, the mould is filled with concrete, covered with a lid and after hardening transferred to its own repository. The deposition in this repository is described. (author)

  16. Remote handling equipment design for the HEDL fuel supply program

    International Nuclear Information System (INIS)

    Metcalf, I.L.

    1984-09-01

    A process line is currently being developed for fabrication of high exposure mixed uranium-plutonium core assemblies. This paper describes the design philosophy, process flow, equipment, and the handling and radiation shielding techniques used for inspection of Fast Flux Test Facility (FFTF) fuel pins and assembly of Driver Fuel Assemblies (DFAs) 6 figures

  17. Fuel elements handling device and method

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1976-01-01

    This invention relates to nuclear equipment and more particularly to methods and apparatus for the non-destructive inspection, manipulation, disassembly and assembly of reactor fuel elements and the like. (author)

  18. The training for nuclear fuel handling at EDF

    International Nuclear Information System (INIS)

    Marion, J.P.

    1999-01-01

    The handling of fuel assemblies in a nuclear power plant presents 3 types of work: the taking delivery of fresh fuel, the refueling and the disposal of spent fuel. These operations are realized by teams made up of 3 handling operators and a supervisor. The refueling is made by 3*8-hour teams. These handling operations are important for the nuclear safety, a mishandling can damage the fuel cladding which is the first containment barrier, so a training center (CETIC) has been created. This center was founded in 1986 by EDF and Framatome, the purpose was to validate maintenance procedures, to test handling equipment and to train the teams which work on site. Various training programmes have been set up and a system of qualification degrees has been organized. The CETIC is fitted up with equipment that are full-sized mockups of real installations. Fuel assemblies don't react in a similar way to the different mechanical and neutronic stresses they undergo while they are in the core, they get deformed and the handling operations become more delicate. The mockup fuel assemblies are quite deformed to train the teams and prepare them to face any real situation. (A.C.)

  19. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Yoon, J. S.; Hong, H. D.; Kim, Y. H.

    2001-03-01

    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

  20. Test plan for K-Basin fuel handling tools

    International Nuclear Information System (INIS)

    Bridges, A.E.

    1995-01-01

    The purpose of this document is to provide the test plan and procedures for the acceptance testing of the handling tools enveloped for the removal of an N-Reactor fuel element from its storage canister in the K-Basins storage pool and insertion into the Single fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N-Reactor fuel elements is part of the overall characterization effort. New hand tools were required since previous fuel movement has involved grasping the fuel in a horizontal position. The 305 Building Cold Test Facility will be used to conduct the acceptance testing of the Fuel Handling Tools. Upon completion of this acceptance testing and any subsequent training of operators, the tools will be transferred to the 105 KW Basin for installation and use

  1. Some factors to consider in handling and storing spent fuel

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1985-11-01

    This report includes information from various studies performed under the Wet Storage Task of the Behavior of Spent Fuel in Storage Project of the Commercial Spent Fuel Management (CSFM) Program at Pacific Northwest Laboratory. Wet storage experience has been summarized earlier in several other reports. This report summarizes pertinent items noted during FY 1985 concerning recent developments in the handling and storage of spent fuel and associated considerations. The subjects discussed include recent publications, findings, and developments associated with: (1) storage of water reactor spent fuel in water pools, (2) extended-burnup fuel, (3) fuel assembly reconstitution and reinsertion, (4) rod consolidation, (5) variations in the US Nuclear Regulatory Commission's definition of failed fuel, (6) detection of failed fuel rods, and (7) extended integrity of spent fuel. A list of pertinent publications is included

  2. Reliability assessment of the fueling machine of the CANDU reactor

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.

    1985-01-01

    Fueling of CANDU-reactors is carried out by two fueling machines, each serving one end of the reactor. The fueling machine becomes a part of the primary heat transport system during the refueling operations, and hence, some refueling machine malfunctions could result in a small scale-loss-of-coolant accident. Fueling machine failures and the failure sequences are discussed. The unavailability of the fueling machine is estimated by using fault tree analysis. The probability of mechanical failure of the fueling machine interface is estimated as 1.08 x 10 -5 . (orig.) [de

  3. West Valley facility spent fuel handling, storage, and shipping experience

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1990-11-01

    The result of a study on handling and shipping experience with spent fuel are described in this report. The study was performed by Pacific Northwest Laboratory (PNL) and was jointly sponsored by the US Department of Energy (DOE) and the Electric Power Research Institute (EPRI). The purpose of the study was to document the experience with handling and shipping of relatively old light-water reactor (LWR) fuel that has been in pool storage at the West Valley facility, which is at the Western New York Nuclear Service Center at West Valley, New York and operated by DOE. A subject of particular interest in the study was the behavior of corrosion product deposits (i.e., crud) deposits on spent LWR fuel after long-term pool storage; some evidence of crud loosening has been observed with fuel that was stored for extended periods at the West Valley facility and at other sites. Conclusions associated with the experience to date with old spent fuel that has been stored at the West Valley facility are presented. The conclusions are drawn from these subject areas: a general overview of the West Valley experience, handling of spent fuel, storing of spent fuel, rod consolidation, shipping of spent fuel, crud loosening, and visual inspection. A list of recommendations is provided. 61 refs., 4 figs., 5 tabs

  4. Method for handling nuclear fuel casks

    International Nuclear Information System (INIS)

    Weems, S.J.

    1976-01-01

    A heavy shielded nuclear fuel cask is lowered into and removed from a water filled spent fuel pool by providing a vertical guide tube in the pool, affixing to the bottom of the cask a base plate that approximates the transverse dimension of the guide tube, and lowering and elevating the cask and base plate assembly into and out of the pool by causing it to traverse within the guide tube. The guide tube and base plate coact to function as a dashpot, thereby cushioning and controlling the fall of the cask in the pool should it break loose while being lowered into or raised out of the pool. a specified approach path to the guide tube insures that the cask assembly will not fall into the pool, should it break loose on its approach to the guide tube

  5. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    Kee, R.W.; Denero, J.V.

    1975-01-01

    An apparatus for loading nuclear fuel pellets on trays for transfer in a system is described. A conveyor supplies pellets from a source to a loading station. When the pellets reach a predetermined position at the loading station, a manual or automatically operated arm pushes the pellets into slots on a tray and this process is repeated until pellet sensing switches detect that the tray is full. Thereupon, the tray is lowered onto a belt or other type conveyor and transferred to other apparatus in the system, such as a furnace for sintering, and in some cases, reduction of UO 2 . 2 to UO 2 . The pellets are retained on the tray and subsequently loaded directly into fuel rods to be used in the reactor core. (auth)

  6. Ontario Hydro Pickering Generating Station fuel handling system performance

    International Nuclear Information System (INIS)

    Underhill, H.J.

    1986-01-01

    The report briefly describes the Pickering Nuclear Generating Station (PNGS) on-power fuel handling system and refuelling cycle. Lifetime performance parameters of the fuelling system are presented, including station incapability charged to the fuel handling system, cost of operating and maintenance, dose expenditure, events causing system unavailability, maintenance and refuelling strategy. It is concluded that the 'CANDU' on-power fuelling system, by consistently contributing less than 1% to the PNGS incapability, has been credited with a 6 to 20% increase in reactor capacity factor, compared to off-power fuelling schemes. (author)

  7. Thimble grip fuel assembly handling tool

    International Nuclear Information System (INIS)

    Salton, R.B.; Hornak, L.P.; Marshall, J.R.; Meuschke, R.E.

    1989-01-01

    This patent describes an apparatus for lifting a fuel assembly of a nuclear reactor. The fuel assembly consists of a top nozzle and control rod guide tubes. The apparatus having a gripping means comprised of: a life plate, an actuating plate having a plurality of apertures, the actuating plate disposed in spaced relationship below the lift plate and vertically movable relative thereto; gripping members operably associated with the lift and actuating plates, the gripping members comprising: (a) a vertical rod fixedly secured near its top end to the lift plate and projecting downward therefrom through an associated aperture in the actuating plate, the rod having a first frustoconical surface formed near its lower end, (b) a generally cylindrical, elastically deformable vertical sleeve having a bore therethrough with a first inner diameter, the sleeve having a first bevelled inside surface near the top end and a second bevelled inside surface at the bottom end of the sleeve, and (c) a vertical gripper actuator disposed about the rod

  8. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    1978-01-01

    In this report the various facilities incorporated in the proposed handling chain for spent fuel from the power stations to the final repository are discribed. Thus the geological conditions which are essential for a final repository is discussed as well as the buffer and canister materials and how they contribute towards a long-term isolation of the spent fuel. Furthermore one chapter deals with leaching of the deposited fuel in the event that the canister is penetrated as well as the transport mechanisms which determine the migration of the radioactive substances through the buffer material. The dispersal processes in the geosphere and the biosphere are also described together with the transfer mechanisms to the ecological systems as well as radiation doses. Finally a summary is given of the safety analysis of the proposed method for the handling and final storage of the spent fuel. (E.R.)

  9. Dry cask handling system for shipping nuclear fuel

    International Nuclear Information System (INIS)

    Jones, C.R.

    1975-01-01

    A nuclear facility is described for improved handling of a shipping cask for nuclear fuel. After being brought into the building, the cask is lowered into a tank mounted on a transporter, which then carries the tank into a position under an auxiliary well to which it is sealed. Fuel can then be loaded into or unloaded from the cask via the auxiliary well which is flooded. Throughout the procedure, the cask surface remains dry. (U.S.)

  10. Computer imaging of EBR-II fuel handling equipment

    International Nuclear Information System (INIS)

    Peters, G.G.; Hansen, L.H.

    1995-01-01

    This paper describes a three-dimensional graphics application used to visualize the positions of remotely operated fuel handling equipment in the EBR-II reactor. A three-dimensional (3D) visualization technique is necessary to simulate direct visual observation of the transfers of fuel and experiments into and out of the reactor because the fuel handling equipment is submerged in liquid sodium and therefore is not visible to the operator. The system described in this paper uses actual signals to drive a three-dimensional computer-generated model in real-time in response to movements of equipment in the plant This paper will present details on how the 3D model of the intank equipment was created and how real-time dynamic behavior was added to each of the moving components

  11. Test reports for K Basins vertical fuel handling tools

    Energy Technology Data Exchange (ETDEWEB)

    Meling, T.A.

    1995-02-01

    The vertical fuel handling tools, for moving N Reactor fuel elements, were tested in the 305 Building Cold Test Facility (CTF) in the 300 Area. After fabrication was complete, the tools were functionally tested in the CTF using simulated N Reactor fuel rods (inner and outer elements). The tools were successful in picking up the simulated N Reactor fuel rods. These tools were also load tested using a 62 pound dummy to test the structural integrity of each assembly. The tools passed each of these tests, based on the performance objectives. Finally, the tools were subjected to an operations acceptance test where K Basins Operations personnel operated the tool to determine its durability and usefulness. Operations personnel were satisfied with the tools. Identified open items included the absence of a float during testing, and documentation required prior to actual use of the tools in the 100 K fuel storage basin.

  12. Development of nuclear fuel microsphere handling techniques and equipment

    International Nuclear Information System (INIS)

    Mack, J.E.; Suchomel, R.R.; Angelini, P.

    1979-01-01

    Considerable progress has been made in the development of microsphere handling techniques and equipment for nuclear applications. Work at Oak Ridge National Laboratory with microspherical fuel forms dates back to the early sixties with the development of the sol-gel process. Since that time a number of equipment items and systems specifically related to microsphere handling and characterization have been identified and developed for eventual application in a remote recycle facility. These include positive and negative pressure transfer systems, samplers, weighers, a blender-dispenser, and automated devices for particle size distribution and crushing strength analysis. The current status of these and other components and systems is discussed

  13. Better fuel handling system performance through improved elastomers and seals

    Energy Technology Data Exchange (ETDEWEB)

    Wensel, R G; Metcalfe, R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1997-12-31

    In the area of elastomers, tests have identified specific compounds that perform well in each class of CANDU service. They offer gains in service life, sometimes by factors of ten or more. Moreover, the aging characteristics of these specific compounds are being thoroughly investigated, whereas many elastomers used previously were either non-specific or their aging was unknown. In this paper the benefits of elastomer upgrading, as well as the deficiencies of current station elastomer practices, are discussed in the context of fuel handling equipment. Guidelines for procurement, storage, handling and condition monitoring of elastomer seals are outlined. (author). 3 figs.

  14. Better fuel handling system performance through improved elastomers and seals

    International Nuclear Information System (INIS)

    Wensel, R.G.; Metcalfe, R.

    1996-01-01

    In the area of elastomers, tests have identified specific compounds that perform well in each class of CANDU service. They offer gains in service life, sometimes by factors of ten or more. Moreover, the aging characteristics of these specific compounds are being thoroughly investigated, whereas many elastomers used previously were either non-specific or their aging was unknown. In this paper the benefits of elastomer upgrading, as well as the deficiencies of current station elastomer practices, are discussed in the context of fuel handling equipment. Guidelines for procurement, storage, handling and condition monitoring of elastomer seals are outlined. (author). 3 figs

  15. Canadian capabilities in fusion fuels technology and remote handling

    International Nuclear Information System (INIS)

    1987-10-01

    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

  16. Man/machine interface for a nuclear cask remote handling control station: system design requirements

    International Nuclear Information System (INIS)

    Clarke, M.M.; Kreifeldt, J.G.; Draper, J.V.

    1984-01-01

    Design requirements are presented for a control station of a proposed semi-automated facility for remote handling of nuclear waste casks. Functional and operational man/machine interface: controls, displays, software format, station architecture, and work environment. In addition, some input is given to the design of remote sensing systems in the cask handling areas. 18 references, 9 figures, 12 tables

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

    International Nuclear Information System (INIS)

    Rowe, J.C.; Spille, R.F.; Zimmermann, S.D.

    1986-12-01

    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

  18. Fuel handling solutions to power pulse at Bruce NGS A

    International Nuclear Information System (INIS)

    Day, R.C.

    1996-01-01

    In response to the discovery of the power pulse problem in March of 1993, Bruce A has installed flow straightening shield plugs in the inner zone channels of all units to partially reduce the gap and gain an increase in reactor power to 75%. After review and evaluation of solutions to manage the gap, including creep compensators and long fuel bundles, efforts have focused on a different solution involving reordering the fuel bundles to reverse the burnup profile. This configuration is maintained by fuelling with the flow and providing better support to the highly irradiated downstream fuel bundles by changing the design of the outlet shield plug. Engineering changes to the fuel handling control system and outlet shield plug are planned to be implemented starting in June 1996, thereby eliminating the power pulse problem and restrictions on reactor operating power. (author). 2 refs., 1 tab., 2 figs

  19. Fuel handling solutions to power pulse at Bruce NGS A

    Energy Technology Data Exchange (ETDEWEB)

    Day, R C [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

    1997-12-31

    In response to the discovery of the power pulse problem in March of 1993, Bruce A has installed flow straightening shield plugs in the inner zone channels of all units to partially reduce the gap and gain an increase in reactor power to 75%. After review and evaluation of solutions to manage the gap, including creep compensators and long fuel bundles, efforts have focused on a different solution involving reordering the fuel bundles to reverse the burnup profile. This configuration is maintained by fuelling with the flow and providing better support to the highly irradiated downstream fuel bundles by changing the design of the outlet shield plug. Engineering changes to the fuel handling control system and outlet shield plug are planned to be implemented starting in June 1996, thereby eliminating the power pulse problem and restrictions on reactor operating power. (author). 2 refs., 1 tab., 2 figs.

  20. Current US strategy and technologies for spent fuel handling

    International Nuclear Information System (INIS)

    Bennett, P.C.; Stringer, J.B.

    1999-01-01

    The United States Department of Energy has recently completed a topical safety analysis report outlining the design and operation of a Centralized Interim Storage Facility for spent commercial nuclear fuel. During the course of the design, dose assessments indicated the need for remote operation of many of the cask handling operations. Use of robotic equipment was identified as a desirable handling solution that is capable of automating many of the operations to maintain throughput, and sufficiently flexible to handle five or more different storage cask designs in varying numbers on a given day. This paper discusses the facility and the dose assessment leading to this choice, and reviews factors to be considered when choosing robotics or automation. Further, a new computer simulation tool to quantify dose to humans working in radiological environments, the Radiological Environment Modeling System (REMS), is introduced. REMS has been developed to produce a more accurate estimate of dose to radiation workers in new activities with radiological hazards. (author)

  1. On current US strategy and technologies for spent fuel handling

    International Nuclear Information System (INIS)

    Bennett, P.C.

    1997-01-01

    The US Department of Energy has recently completed a topical safety analysis report outlining the design and operation of a Centralized Interim Storage Facility for spent commercial nuclear fuel. During the course of the design, dose assessments indicated the need for remote operation of many of the cask handling operations. Use of robotic equipment was identified as a desirable handling solution that is capable of automating many of the operations to maintain throughput, and sufficiently flexible to handle five or more different storage cask designs in varying numbers on a given day. This paper discusses the facility and the dose assessment leading to this choice, and reviews factors to be considered when choosing robotics or automation. Further, a new computer simulation tool to quantify dose to humans working in radiological environments, the Radiological Environment Modeling System (REMS), is introduced. REMS has been developed to produce a more accurate estimate of dose to radiation workers in new activities with radiological hazards

  2. Evolution of the Darlington NGS fuel handling computer systems

    International Nuclear Information System (INIS)

    Leung, V.; Crouse, B.

    1996-01-01

    The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety. Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment. Combining modern technology upgrades into a system of obsolete hardware components is not an easy task. Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g. information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. (author). 3 figs

  3. Feasibility study of CANDU-9 fuel handling system

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Jeong Ki; Jo, C. H.; Kim, H. M.

    1996-12-01

    CANDU`s combination of natural uranium, heavy water and on-power refuelling is unique in its design and remarkable for reliable power production. In order to offer more output, better site utilization, shorter construction time, improved station layout, safety enhancements and better control panel layout, CANDU-9 is now under development with design improvement added to all proven CANDU advantages or applicable technologies. One of its major improvement has been applied to fuel handling system. This system is very similar to that of CANDU-3, and some parts of the system are applied to those of the existing CANDU-6 or CANDU-9. Design concepts and design requirements of fuel handling system for CANDU-9 have been identified to compare with those of the existing CANDU and the design feasibilities have been evaluated. (author). 4 tabs., 13 figs., 9 refs.

  4. EBR-II fuel handling console digital upgrade

    International Nuclear Information System (INIS)

    Peters, G.G.; Wiege, D.D.; Christensen, L.J.

    1995-01-01

    The main fuel handling console and control system at the Experimental Breeder Reactor II (EBR-II) are being upgraded to a computerized system using high-end workstations for the operator interface and a programmable logic controller (PLC) for the control system. Two-dimensional (2D) and three-dimensional (3D) computer graphics will be provided for the operator which will show the relative position of under-sodium fuel handling equipment. This equipment is operated remotely with no means of directly viewing the transfer. This paper describes various aspects of the modification including reasons for the upgrade, capabilities the new system provides over the old control system, philosophies and rationale behind the new design, testing and simulation work, diagnostic features, and the advanced graphics techniques used to display information to the operator

  5. Evolution of the Darlington NGS fuel handling computer systems

    Energy Technology Data Exchange (ETDEWEB)

    Leung, V; Crouse, B [Ontario Hydro, Bowmanville (Canada). Darlington Nuclear Generating Station

    1997-12-31

    The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety. Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment. Combining modern technology upgrades into a system of obsolete hardware components is not an easy task. Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g. information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. (author). 3 figs.

  6. Cellular Handling of Protein Aggregates by Disaggregation Machines.

    Science.gov (United States)

    Mogk, Axel; Bukau, Bernd; Kampinga, Harm H

    2018-01-18

    Both acute proteotoxic stresses that unfold proteins and expression of disease-causing mutant proteins that expose aggregation-prone regions can promote protein aggregation. Protein aggregates can interfere with cellular processes and deplete factors crucial for protein homeostasis. To cope with these challenges, cells are equipped with diverse folding and degradation activities to rescue or eliminate aggregated proteins. Here, we review the different chaperone disaggregation machines and their mechanisms of action. In all these machines, the coating of protein aggregates by Hsp70 chaperones represents the conserved, initializing step. In bacteria, fungi, and plants, Hsp70 recruits and activates Hsp100 disaggregases to extract aggregated proteins. In the cytosol of metazoa, Hsp70 is empowered by a specific cast of J-protein and Hsp110 co-chaperones allowing for standalone disaggregation activity. Both types of disaggregation machines are supported by small Hsps that sequester misfolded proteins. Copyright © 2018 Elsevier Inc. All rights reserved.

  7. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

    Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

  8. Handling final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    1978-01-01

    The present second report from KBS describes how the safe final storage of spent unreprocessed nuclear fuel can be implemented. According to the Swedish Stipulation Law, the owner must specify in which form the waste is to be stored, how final storage is to be effected, how the waste is to be transported and all other aspects of fuel handling and storage which must be taken into consideration in judging whether the proposed final storage method can be considered to be absolutely safe and feasible. Thus, the description must go beyond general plans and sketches. The description is therefore relatively detailed, even concerning those parts which are less essential for evaluating the safety of the waste storage method. For those parts of the handling chain which are the same for both alternatives of the Stipulation Law, the reader is referred in some cases to the first report. Both of the alternatives of the Stipulation Law may be used in the future. Handling equipment and facilities for the two storage methods are so designed that a combination in the desired proportions is practically feasible. In this first part of the report are presented: premises and data, a description of the various steps of the handling procedure, a summary of dispersal processes and a safety analysis. (author)

  9. Handling encapsulated spent fuel in a geologic repository environment

    International Nuclear Information System (INIS)

    Ballou, L.B.

    1983-02-01

    In support of the Spent Fuel Test-Climate at the U.S. Department of Energy's Nevada Test Site, a spent-fuel canister handling system has been designed, deployed, and operated successfully during the past five years. This system transports encapsulated commercial spent-fuel assemblies between the packaging facility and the test site (approx. 100 km), transfers the canisters 420 m vertically to and from a geologic storage drift, and emplaces or retrieves the canisters from the storage holes in the floor of the drift. The spent-fuel canisters are maintained in a fully shielded configuration at all times during the handling cycle, permitting manned access at any time for response to any abnormal conditions. All normal operations are conducted by remote control, thus assuring as low as reasonably achievable exposures to operators; specifically, we have had no measurable exposure during 30 canister transfer operations. While not intended to be prototypical of repository handling operations, the system embodies a number of concepts, now demonstrated to be safe, reliable, and economical, which may be very useful in evaluating full-scale repository handling alternatives in the future. Among the potentially significant concepts are: Use of an integral shielding plug to minimize radiation streaming at all transfer interfaces. Hydraulically actuated transfer cask jacking and rotation features to reduce excavation headroom requirements. Use of a dedicated small diameter (0.5 m) drilled shaft for transfer between the surface and repository workings. A wire-line hoisting system with positive emergency braking device which travels with the load. Remotely activated grapples - three used in the system - which are insensitive to load orientation. Rail-mounted underground transfer vehicle operated with no personnel underground

  10. Baseline descriptions for LWR spent fuel storage, handling, and transportation

    International Nuclear Information System (INIS)

    Moyer, J.W.; Sonnier, C.S.

    1978-04-01

    Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables

  11. Baseline descriptions for LWR spent fuel storage, handling, and transportation

    Energy Technology Data Exchange (ETDEWEB)

    Moyer, J.W.; Sonnier, C.S.

    1978-04-01

    Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables.

  12. Nuclear fuel handling grapple carriage with self-lubricating bearing

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    Disclosed is a nuclear fuel handling grapple carriage having a bearing with a lubricant reservoir that is capable of being refilled when the bearing and reservoir are submerged in a lubricant pool. The lubricant reservoir supplies lubricant to the bearing while the bearing allows a small amount of lubricant to leak passed appropriately placed seals creating a positive out flow of lubricant thereby preventing foreign material from entering the bearing

  13. Progress in control equipment for fuel-handling machinery

    International Nuclear Information System (INIS)

    Nutting, B.A.

    1986-01-01

    The paper outlines the development of the equipment used to control the fuel-handling machinery associated with nuclear reactors, from the early electromechanical equipment, through solid-state switching logic to programmable controllers and microprocessors. The control techniques have developed along with the technology, and modern systems offer versatility, reliability and ease of design, operation and maintenance. Future trends and developments are discussed together with possible limiting factors. (author)

  14. Interim report spent nuclear fuel retrieval system fuel handling development testing

    Energy Technology Data Exchange (ETDEWEB)

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  15. Handling of spent fuel from research reactors in Japan

    International Nuclear Information System (INIS)

    Kanda, K.

    1997-01-01

    In Japan eleven research reactors are in operation. After the 19th International Meeting on Reduced Enrichment for Research Reactors and Test Reactors (RERTR) on October 6-10, 1996, Seoul, Korea, the Five Agency Committee on Highly Enriched Uranium, which consists of Science and Technology Agency, the Ministry of Education, Science and Culture, the Ministry of Foreign Affairs, Japan Atomic Energy Research Institute (JAERI) and Kyoto University Research Reactor Institute (KURRI) met on November 7,1996, to discuss the handling of spent fuel from research reactors in Japan. Advantages and disadvantages to return spent fuel to the USA in comparison to Europe were discussed. So far, a number of spent fuel elements in JAERI and KURRI are to be returned to the US. The first shipment to the US is planned for 60 HEU elements from JMTR in 1997. The shipment from KURRI is planned to start in 1999. (author)

  16. Pacific Northwest Laboratory (PNL) spent fuel transportation and handling facility models

    International Nuclear Information System (INIS)

    Andrews, W.B.; Bower, J.C.; Burnett, R.A.; Engel, R.L.; Rolland, C.W.

    1979-09-01

    A spent fuel logistics study was conducted in support of the US DOE program to develop facilities for preparing spent unreprocessed fuel from commercial LWRs for geological storage. Two computerized logistics models were developed. The first one was the site evaluation model. Two studies of spent fuel handling facility and spent fuel disposal facility siting were completed; the first postulates a single spent fuel handling facility located at any of six DOE laboratory sites, while the second study examined siting strategies with the spent fuel repository relative to the spent fuel handling facility. A second model to conduct storage/handling facility simulations was developed

  17. Pacific Northwest Laboratory (PNL) spent fuel transportation and handling facility models

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, W.B.; Bower, J.C.; Burnett, R.A.; Engel, R.L.; Rolland, C.W.

    1979-09-01

    A spent fuel logistics study was conducted in support of the US DOE program to develop facilities for preparing spent unreprocessed fuel from commercial LWRs for geological storage. Two computerized logistics models were developed. The first one was the site evaluation model. Two studies of spent fuel handling facility and spent fuel disposal facility siting were completed; the first postulates a single spent fuel handling facility located at any of six DOE laboratory sites, while the second study examined siting strategies with the spent fuel repository relative to the spent fuel handling facility. A second model to conduct storage/handling facility simulations was developed. (DLC)

  18. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    International Nuclear Information System (INIS)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems' Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment

  19. A positive action handling tool for TRIGA fuel

    International Nuclear Information System (INIS)

    McMaster, Ira B.

    1976-01-01

    Because several elements have disengaged accidentally from the conventional fuel handling tool at the PSBR a need was apparent for a tool whose action was more positive. The new design utilizes rotary motion to provide a positive locking action when the tool engages an element. This action provides a secure grip on the element and positive control by the tool operator over when an element can disengage from the tool. The convenience provided by the flexibility of the original tool is retained by making the lower four feet of the new tool flexible. (author)

  20. Sodium removal from the grapples of the fuel handling facility of Joyo

    Energy Technology Data Exchange (ETDEWEB)

    Mukaibo, R; Matsuno, Y; Sato, I; Yoneda, Y; Sato, H [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

    1978-08-01

    Sodium removal from the grapples of the fuel handling facility of 'JOYO' is done in alcohol. The operations of the cleaning facility started as the functional tests of the fuel handling facility began. Since then, criticality test and low power tests had been done and during this period, sodium removal from the grapples, after a certain amount of time in use, were done. In order to lessen the time for the cleaning process for the grapples of the machines inside the containment vessel, demineralized water concentration in the alcohol was gained to as much as 10% and good results were obtained. On the other hand, there were very small amounts of sodium on the grapples of the machine used outside the containment vessel and direct charging of demineralized water into the cleaning pot was done experimentally, also with good results. In this report, the sodium removal experience of the grapples before power up tests and some remarks on the improvements of the facility for the future are presented. (author)

  1. Operator aid system for Dhruva fueling machine

    International Nuclear Information System (INIS)

    Misra, S.M.; Ramaswamy, L.R.; Gohel, N.; Bharadwaj, G.; Ranade, M.R.; Khadilkar, M.G.

    1997-01-01

    Systems with significant software contents are replacing the old hardware logic systems. These systems not only are versatile but are easy to make changes in the program. Extensive use of such systems in critical real-time operation environment warrants not only excessive training on simulators, documentation but also fault tolerant system to bring the operation to a safe state in case of error. With new graphic user software interface and advancement in personal computer hardware design, the dynamic status of the physical environment can be shown on the visual display at near real time. These visual aids along with the software covering all the interlocks aids an operator in his professional work. This paper highlights the operator aid system for Dhruva fueling machine. (author). 6 refs., 1 fig

  2. Fort St. Vrain fuel-handling system RAM analysis

    International Nuclear Information System (INIS)

    Azizi, S.M.; Berg, G.E.; Burton, J.H.; Durand, R.E.; Larson, E.M.; Pepe, D.J.; Rutherford, P.D.; Novachek, F.J.

    1989-01-01

    Public Service of Company of Colorado (PSC) is planning to decommission its Fort St. Vrain plant in 1990. This requires removal of 1,500 separate assemblies from the core. With the low historical availability of the fuel-handling system (FHS), defueling time was estimated at 36 months. With plant expenses of approximately $1.6 million per month during defueling, this would mean a schedule cost of $58 million. With their contractor, Rockwell International, PSC embarked on a reliability, availability, and maintainability (RAM) analysis to reduce projected defueling time. Key elements included (a) estimating availability of the FHS using a limited historical record, (b) assessing the defueling critical path, and (c) proposing and evaluating design/operational improvements. The most cost-effective improvements are being implemented and are expected to provide a reduction of >18 months in schedule and a net savings of $20 to 25 million. The paper describes the FHS design and operation, major problems associated with fuel-handling operations, and results and recommendations

  3. Advanced operator interface design for CANDU-3 fuel handling system

    International Nuclear Information System (INIS)

    Arapakota, D.

    1995-01-01

    The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System'. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author)

  4. FUEL HANDLING FACILITY BACKUP CENTRAL COMMUNICATIONS ROOM SPACE REQUIREMENTS CALCULATION

    International Nuclear Information System (INIS)

    SZALEWSKI, B.

    2005-01-01

    The purpose of the Fuel Handling Facility Backup Central Communications Room Space Requirements Calculation is to determine a preliminary estimate of the space required to house the backup central communications room in the Fuel Handling Facility (FHF). This room provides backup communications capability to the primary communication systems located in the Central Control Center Facility. This calculation will help guide FHF designers in allocating adequate space for communications system equipment in the FHF. This is a preliminary calculation determining preliminary estimates based on the assumptions listed in Section 4. As such, there are currently no limitations on the use of this preliminary calculation. The calculations contained in this document were developed by Design and Engineering and are intended solely for the use of Design and Engineering in its work regarding the FHF Backup Central Communications Room Space Requirements. Yucca Mountain Project personnel from Design and Engineering should be consulted before the use of the calculations for purposes other than those stated herein or use by individuals other than authorized personnel in Design and Engineering

  5. Advanced operator interface design for CANDU-3 fuel handling system

    Energy Technology Data Exchange (ETDEWEB)

    Arapakota, D [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System`. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author).

  6. Remote handling and automation in back end of fuel cycle

    International Nuclear Information System (INIS)

    Nair, K.N.S.

    2010-01-01

    Full text: Indian nuclear programme is readying for a quantum leap and it is essential that technology is available for building advanced fuel recycle plants in the back end and for sustained operation of such plants. Remote technology and automation plays a big role to achieve this goal. With the introduction of advanced fuel cycles in indigenous programme and scenario of international cooperation it is essential to be ready with indigenous technology for meeting all challenges. Work has been progressing to develop locally support technology for remote handling and automation with good success. Essential RH tools such as master slave manipulators, power manipulators and hot cell viewing systems have been developed and commercial production has been established. Customised RH requirements for back end plants have been met and the designs have proven to be worthy for hot operations over the years. In the last few years stress has been on development of equipment and technology to meet the increasing demands of higher throughput plants. Substantial progress has been achieved in the head end and reconversion laboratory systems of reprocessing plants. Similarly successful efforts have also been made for establishing Thoria processing cells and also the RH in the reconversion operations. Custom designed equipment has been developed for decommissioning of ceramic melter, used glove boxes etc. Efforts are on hand to develop automated RH equipment for material handling in underground repositories. This paper aims at bringing out the theme based on some of our own experiences and some reports from plants in operation abroad. (author)

  7. Fuel handling system of Indian 500 MWe PHWR-evolution and innovations

    International Nuclear Information System (INIS)

    Sanatkumar, A.; Jit, I.; Muralidhar, G.

    1996-01-01

    India has gained rich experience in design, manufacture, testing, operation and maintenance of the Fuel Handling System of CANDU type PHWRs. When design and layout of the first 500 MWe PHWR was being evolved, it was possible for us to introduce many special and innovative features in the Fuel Handling System which are friendly for operations and maintenance personnel. Some of these are: Simple, robust and modular mechanisms for ease of maintenance; Shorter turnaround time for refuelling a channel by introduction of transit equipment between the Fuelling Machine (FM) Head and light water equipment; Optimised layout to transport spent fuel in straight and short path and also to facilitate direct wheeling out of the FM Head from the Reactor Building to the Service Building; Provision to operate the FM Head even when the Primary Heat Transport (PHT) System is open for maintenance; Control-console engineered for carrying out refuelling operations in the sitting position; and, Dedicated calibration and maintenance facility to facilitate quick replacement of the FM Head as a single unit. The above special features have been described in this paper. (author). 7 figs

  8. Fuel handling system of Indian 500 MWe PHWR-evolution and innovations

    Energy Technology Data Exchange (ETDEWEB)

    Sanatkumar, A; Jit, I; Muralidhar, G [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    India has gained rich experience in design, manufacture, testing, operation and maintenance of the Fuel Handling System of CANDU type PHWRs. When design and layout of the first 500 MWe PHWR was being evolved, it was possible for us to introduce many special and innovative features in the Fuel Handling System which are friendly for operations and maintenance personnel. Some of these are: Simple, robust and modular mechanisms for ease of maintenance; Shorter turnaround time for refuelling a channel by introduction of transit equipment between the Fuelling Machine (FM) Head and light water equipment; Optimised layout to transport spent fuel in straight and short path and also to facilitate direct wheeling out of the FM Head from the Reactor Building to the Service Building; Provision to operate the FM Head even when the Primary Heat Transport (PHT) System is open for maintenance; Control-console engineered for carrying out refuelling operations in the sitting position; and, Dedicated calibration and maintenance facility to facilitate quick replacement of the FM Head as a single unit. The above special features have been described in this paper. (author). 7 figs.

  9. Considerations for handling failed fuel at the Barnwell Nuclear Fuel Plant

    International Nuclear Information System (INIS)

    Anderson, R.T.; Cholister, R.J.

    1982-05-01

    The impact of failed fuel receipt on reprocessing operations is qualitatively described. It appears that extended storage of fuel, particularly with advanced storage techniques, will increase the quantity of failed fuel, the nature and possibly the configuration of the fuel. The receipt of failed fuel at the BNFP increases handling problems, waste volumes, and operator exposure. If it is necessary to impose special operating precautions to minimize this impact, a loss in plant throughput will result. Hence, ideally, the reprocessing plant operator would take every reasonable precaution so that no failed fuel is received. An alternative policy would be to require that failed fuel be placed in a sealed canister. In the latter case the canister must be compatible with the shipping cask and suitable for in-plant storage. A required inspection of bare fuel would be made at the reactor prior to shipping off-site. This would verify fuel integrity. These requirements are obviously idealistic. Due to the current uncertain status of reprocessing and the need to keep reactors operating, business or governmental policy may be enacted resulting in the receipt of a negotiated quantity of non-standard fuel (including failed fuel). In this situation, BNFP fuel receiving policy based soley on fuel cladding integrity would be difficult to enforce. There are certain areas where process incompatibility does exist and where a compromise would be virtually impossible, e.g., canned fuel for which material or dimensional conflicts exist. This fuel would have to be refused or the fuel would require recanning prior to shipment. In other cases, knowledge of the type and nature of the failure may be acceptable to the operator. A physical inspection of the fuel either before shipment or after the cask unloading operation would be warranted. In this manner, concerns with pool contamination can be identified and the assembly canned if deemed necessary

  10. Soil-structure interaction in fuel handling building

    International Nuclear Information System (INIS)

    Elaidi, B.M.; Eissa, M.A.

    1998-01-01

    This paper presents an accurate three-dimensional seismic soil-structure interaction analysis for large structures. The method is applied to the fuel building in nuclear power plants. The analysis is performed numerically in the frequency domain and the responses are obtained by inverse Fourier transformation. The size of the structure matrices is reduced by transforming the equation of motion to the modal coordinate system. The soil is simulated as a layered media on top of viscoelastic half space. Soil impedance matrices are calculated from the principles of continuum mechanics and account for soil stiffness and energy dissipation. Effects of embedment on the field equations is incorporated through the scattering matrices or by simply scaling the soil impedance. Finite element methods are used to discretize the concrete foundation for the generation of the soil interaction matrices. Decoupling of the sloshing water in the spent fuel pools and the free-standing spent fuel racks is simulated. The input seismic motions are defined by three artificial time history accelerations. These input motions are generated to match the ground design basis response spectra and the target power spectral density function. The methods described in this paper can handle arbitrary foundation layouts, allows for large structural models, and accurately represents the soil impedance. Time history acceleration responses were subsequently used to generate floor response spectra at applicable damping values. (orig.)

  11. Machine assembly with a new material handling mechanism in the sewing machine

    Directory of Open Access Journals (Sweden)

    Umarova Z.M.

    2017-05-01

    Full Text Available the paper presents the dynamic model of the machine assembly with a recommended mechanism for moving material and the definition of the law of rails motion under various system parameters. The author has suggested the solution implemented by the system of differential equations numerically on the PC and the system describing the motion of the machine set. Recommended values ​​of the parameters of elastic links of material transfer mechanism have been obtained. The researcher has developed the methods of kinematic and dynamic analysis of the material transfer mechanism with elastic elements of the sewing machine and has approved the parameters and development of the design.

  12. Simulator for candu600 fuel handling system. the experimental model

    International Nuclear Information System (INIS)

    Marinescu, N.; Predescu, D.; Valeca, S.

    2013-01-01

    A main way to increase the nuclear plant safety is related to selection and continuous training of the operation staff. In this order, the computer programs for training, testing and evaluation of the knowledge get, or training simulators including the advanced analytical models of the technological systems are using. The Institute for Nuclear Research from Pitesti, Romania intend to design and build an Fuel Handling Simulator at his F/M Head Test Rig facility, that will be used for training of operating personnel. This paper presents simulated system, advantages to use the simulator, and the experimental model of simulator, that has been built to allows setting of the requirements and fabrication details, especially for the software kit that will be designed and implement on main simulator. (authors)

  13. Fuel Handling Equipment Maintenance for Critical Path Time Savings

    Energy Technology Data Exchange (ETDEWEB)

    Saville, M.; Williams, A.

    2015-07-01

    By sharing lessons learned and operating experience gained by AREVA Stearns RogerTM Services from more than 45 years of servicing, maintaining, and upgrading Fuel Handling Equipment (FHE) and as the original equipment manufacturer to 56% of domestic U.S. FHE (PWR and BWR) as well as 19 units overseas, this paper presents trends and market forces that have led to the neglect of FHE, the risks of not adequately maintaining FHE, and the financial benefits of proactively maintaining FHE. The benefit to audiences is to come to a better understanding of how critical path delays can be avoided and thus reduce nuclear power plant operating costs. Note that statistics and monetary values given herein are based on recent typical experiences of AREVA Stearns RogerTM Services. Examples discussed are based on actual lessons learned. For the purposes of this paper, upgrades are considered a part of equipment maintenance unless specifically discussed separately. (Author)

  14. Development of nuclear fuel cycle remote handling technology

    International Nuclear Information System (INIS)

    Kim, K. H.; Park, B. S.; Kim, S. H.

    2012-04-01

    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

  15. Darlington NGD fuel handling head eight acceptance program

    International Nuclear Information System (INIS)

    Skelton, P.H.; Sie, T.

    1996-01-01

    Darlington NGD requires eight fuelling machine heads to fuel the four 932 MW reactors. Six heads are used on the three fuelling machine trolleys for normal fuelling operations. A further two heads are required to allow for maintenance and to provide for such reactor face activities as PIPE and CIGAR. Seven heads were successfully delivered to site from the head supplier. During acceptance testing, stalls on the charge tube screw assembly of the eighth and final head prevented its delivery to site. Replacement of the charge tube screw with a spare screw did not alleviate the problem. An in depth series of tests were undertaken at site, at the supplier and at the screw sub-supplier to determine the root cause of the problem. These tests included taking torque measurements under different operating conditions and using different components to assess the effects of the changes on torque levels. An assessment of the effects of changing chemical conditions (particularly crud levels) was also made. To ensure that the results of the testing were well understood, additional torque testing was also completed on a head and screw assembly at site that was known to work well. Based on all of the above series of tests, a recommendation was made to re-machine the charge tube screw(s). The original charge tube screw from Head eight was subsequently returned to the sub-supplier for re-work. Follow-up torque measurements and acceptance testing showed that the screw rework was effective and that Head eight could be successfully delivered to site. This paper focuses on the results of the head/screw test program. Results of the acceptance testing are also discussed. (author). 2 refs., 4 figs

  16. Darlington NGD fuel handling head eight acceptance program

    Energy Technology Data Exchange (ETDEWEB)

    Skelton, P H; Sie, T [Ontario Hydro, Bowmanville (Canada). Darlington Nuclear Generating Station; Pilgrim, J [Canadian General Electric Co. Ltd., Toronto, ON (Canada)

    1997-12-31

    Darlington NGD requires eight fuelling machine heads to fuel the four 932 MW reactors. Six heads are used on the three fuelling machine trolleys for normal fuelling operations. A further two heads are required to allow for maintenance and to provide for such reactor face activities as PIPE and CIGAR. Seven heads were successfully delivered to site from the head supplier. During acceptance testing, stalls on the charge tube screw assembly of the eighth and final head prevented its delivery to site. Replacement of the charge tube screw with a spare screw did not alleviate the problem. An in depth series of tests were undertaken at site, at the supplier and at the screw sub-supplier to determine the root cause of the problem. These tests included taking torque measurements under different operating conditions and using different components to assess the effects of the changes on torque levels. An assessment of the effects of changing chemical conditions (particularly crud levels) was also made. To ensure that the results of the testing were well understood, additional torque testing was also completed on a head and screw assembly at site that was known to work well. Based on all of the above series of tests, a recommendation was made to re-machine the charge tube screw(s). The original charge tube screw from Head eight was subsequently returned to the sub-supplier for re-work. Follow-up torque measurements and acceptance testing showed that the screw rework was effective and that Head eight could be successfully delivered to site. This paper focuses on the results of the head/screw test program. Results of the acceptance testing are also discussed. (author). 2 refs., 4 figs.

  17. Radioactive waste management decommissioning spent fuel storage. V. 3. Waste transport, handling and disposal spent fuel storage

    International Nuclear Information System (INIS)

    1985-01-01

    As part of the book entitled Radioactive waste management decommissioning spent fuel storage, vol. 3 dealts with waste transport, handling and disposal, spent fuel storage. Twelve articles are presented concerning the industrial aspects of nuclear waste management in France [fr

  18. Multi Canister Overpack (MCO) Handling Machine - Independent Review of Seismic Structural Analysis

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    The following separate reports and correspondence pertains to the independent review of the seismic analysis. The original analysis was performed by GEC-Alsthom Engineering Systems Limited (GEC-ESL) under subcontract to Foster-Wheeler Environmental Corporation (FWEC) who was the prime integration contractor to the Spent Nuclear Fuel Project for the Multi-Canister Overpack (MCO) Handling Machine (MHM). The original analysis was performed to the Design Basis Earthquake (DBE) response spectra using 5% damping as required in specification, HNF-S-0468 for the 90% Design Report in June 1997. The independent review was performed by Fluor-Daniel (Irvine) under a separate task from their scope as Architect-Engineer of the Canister Storage Building (CSB) in 1997. The comments were issued in April 1998. Later in 1997, the response spectra of the Canister Storage Building (CSB) was revised according to a new soil-structure interaction analysis and accordingly revised the response spectra for the MHM and utilized 7% damping in accordance with American Society of Mechanical Engineers (ASME) NOG-1, ''Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).'' The analysis was re-performed to check critical areas but because manufacturing was underway, designs were not altered unless necessary. FWEC responded to SNF Project correspondence on the review comments in two separate letters enclosed. The dispositions were reviewed and accepted. Attached are supplier source surveillance reports on the procedures and process by the engineering group performing the analysis and structural design. All calculation and analysis results are contained in the MHM Final Design Report which is part of the Vendor Information File 50100. Subsequent to the MHM supplier engineering analysis, there was a separate analyses for nuclear safety accident concerns that used the electronic input data files provided by FWEC/GEC-ESL and are contained in document SNF-6248

  19. Fuel handling alternatives to prepare for large scale fuel channel replacement

    International Nuclear Information System (INIS)

    Martire, S.; Sandu, I.

    2007-01-01

    It is desirable to reduce the duration of defuelling the reactor in preparation for retube, as the cost of replacement power is $750K/day. Three fast defuelling concepts are presented. With the Through Flow Defuelling method, the fuel string is hydraulically pushed into the downstream Fuelling Machine (FM) by flow passing through the fuel channel. The Long Stroke C Ram method replaces the FM C Ram with a longer one capable of pushing all fuel bundles into the receiving FM. Defuelling Hardware uses enhanced design of ram extensions that interconnect mechanically to extend the Ram stroke to push fuel bundles into the receiving FM. This paper will present descriptions of each defuelling concept to prepare for Large Scale Fuel Channel Replacement. Advantages and disadvantages of each concept will be discussed and a recommendation will be made for future implementation. (author)

  20. Mitigating fuel handling situations during station blackout in TAPP-3 and

    International Nuclear Information System (INIS)

    Chugh, V.K.; Roy, Shibaji; Gupta, H.; Inder Jit

    2002-01-01

    Full text: On power refueling is one of the important features of PHWRs. fuelling machine (FM) Head becomes part of the reactor pressure boundary during refueling operations. Hot irradiated (spent) fuel bundles are received in the FM Head from the Reactor and transferred to spent fuel storage bay (SFSB). These bundles pass through various fuel handling (FH) Equipment under submerged condition except during the dry transfer operation. Situations of station blackout (SBO) i.e. postulated simultaneous failure of Class III and Class IV electric power, could persist for a long period, during on-reactor or off-reactor FH operations, with the spent fuel bundles being any where in the system between the reactor and SFSB. The cooling provisions for the spent fuel bundles vary depending upon the stage of operation. During SBO, it becomes difficult to maintain cooling to these fuel bundles due to the limited availability of Class II power and instrument air. However, cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like stay-put, gravity- fill, D 2 O-steaming etc. for cooling the bundles. Various scenarios have been identified for cooling provisions of the bundles in the system. The paper also describes consequences like loss of D 2 O inventory, rise in ambient temperature and pressure and tritium build-up in Reactor Building, emanating from these cooling schemes

  1. FFTF [Fast Flux Test Facility] fuel handling experience (1979--1986)

    International Nuclear Information System (INIS)

    Romrell, D.M.; Art, D.M.; Redekopp, R.D.; Waldo, J.B.

    1987-05-01

    The Fast Flux Test Facility (FFTF)is a 400 MW (th) sodium-cooled fast flux test reactor located on the Hanford Site in southeastern Washington State. The FFTF is operated by the Westinghouse Hanford Company for the United States Department of Energy. The FFTF is a three loop plant designed primarily for the purpose of testing full-scale core components in an environment prototypic of future liquid metal reactors. The plant design emphasizes features to enhance this test capability, especially in the area of the core, reactor vessel, and refueling system. Eight special test positions are provided in the vessel head to permit contact instrumented experiments to be installed and irradiated. These test positions effectively divide the core into three sectors. Each sector requires its own In-Vessel Handling Machine (IVHM) to access all the core positions. Since the core and the in-vessel refueling components are submerged under sodium, all handling operations must be performed blind. This puts severe requirements on the positioning ability are reliability of the refueling components. This report addresses the operating experience with the fuel handling system from initial core loading in November, 1979 through 1986. This includes 9 refueling cycles. 2 refs., 8 figs

  2. Experience with failed or damaged spent fuel and its impacts on handling

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1989-12-01

    Spent fuel management planning needs to include consideration of failed or damaged spent light-water reactor (LWR) fuel. Described in this paper, which was prepared under the Commercial Spent Fuel Management (CSFM) Program that is sponsored by the US Department of Energy (DOE), are the following: the importance of fuel integrity and the behavior of failed fuel, the quantity and burnup of failed or damaged fuel in storage, types of defects, difficulties in evaluating data on failed or damaged fuel, experience with wet storage, experience with dry storage, handling of failed or damaged fuel, transporting of fuel, experience with higher burnup fuel, and conclusions. 15 refs

  3. Proceedings of the 5th international conference on stability and handling of liquid fuels. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Giles, H.N. [ed.

    1995-03-01

    This proceedings summarizes recent work concerning stability and handling of fuels. Jet fuels, gasolines, heavy oils, and distillate fuels were considered. Fuel issues relevant to environmental mandates were discussed. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  4. Spent fuel handling and storage facility for an LWR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Baker, W.H.; King, F.D.

    1979-01-01

    The facility will have the capability to handle spent fuel assemblies containing 10 MTHM/day, with 30% if the fuel received in legal weight truck (LWT) casks and the remaining fuel received in rail casks. The storage capacity will be about 30% of the annual throughput of the reprocessing plant. This size will provide space for a working inventory of about 50 days plant throughput and empty storage space to receive any fuel that might be in transit of the reprocessing plant should have an outage. Spent LWR fuel assemblies outside the confines of the shipping cask will be handled and stored underwater. To permit drainage, each water pool will be designed so that it can be isolated from the remaining pools. Pool water quality will be controlled by a filter-deionizer system. Radioactivity in the water will be maintained at less than or equal to 2 x 10 -4 Ci/m 3 ; conductivity will be maintained at 1 to 2 μmho/cm. The temperature of the pool water will be maintained at less than or equal to 40 0 C to retard algae growth and reduce evaporation. Decay heat will be transferred to the environment via a heat exchanger-cooling tower system

  5. Cleaning, disassembly, and requalification of the FFTF in vessel handling machine

    International Nuclear Information System (INIS)

    Coops, W.J.

    1977-10-01

    The Engineering Model In Vessel Handling Machine (IVHM) was successfully removed, cleaned, disassembled, inspected, reassembled and reinstalled into the sodium test vessel at Richland, Washington. This was the first time in the United States a full size operational sodium wetted machine has been cleaned by the water vapor nitrogen process and requalified for operation. The work utilized an atmospheric control system during removal, a tank type water vapor nitrogen cleaning system and an open ''hands on'' disassembly and assembly stand. Results of the work indicate the tools, process and equipment are adequate for the non-radioactive maintenance sequence. Additionally, the work proves that a machine of this complexity can be successfully cleaned, maintained and re-used without the need to replace a large percentage of the sodium wetted parts

  6. Performance Evaluation and Suggestion of Upgraded Fuel Handling Equipment for Operating OPR1000

    International Nuclear Information System (INIS)

    Chang, Sang Gyoon; Hwang, Jeung Ki; Choi, Taek Sang; Na, Eun Seok; Lee, Myung Lyul; Baek, Seung Jin; Kim, Man Su; Kunik, Jack

    2011-01-01

    The purpose of this study is to evaluate the performance of upgraded FHE (Fuel Handling Equipment) for operating OPR 1000 (Optimized Power Reactor) by using data measured during the fuel reloading, and make some suggestions on enhancing the performance of the FHE. The fuel handling equipment, which serves critical processes in the refueling outage, has been upgraded to increase and improve the operational availability of the plant. The evaluation and suggestion of this study can be a beneficial tool related to the performance of the fuel handling equipment

  7. Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

    Energy Technology Data Exchange (ETDEWEB)

    Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

    1984-01-01

    The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

  8. Nuclear fuel replacement device

    International Nuclear Information System (INIS)

    Ritz, W.C.; Robey, R.M.; Wett, J.F.

    1984-01-01

    A fuel handling arrangement for a liquid metal cooled nuclear reactor having a single rotating plug eccentric to the fuel core and a fuel handling machine radially movable along a slot in the plug with a transfer station disposed outside the fuel core but covered by the eccentric plug and within range of movement of said fuel handling machine to permit transfer of fuel assemblies between the core and the transfer station. (author)

  9. Storage, handling and movement of fuel and related components at nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    The report describes in general terms the various operations involved in the handling of fresh fuel, irradiated fuel, and core components such as control rods, neutron sources, burnable poisons and removable instruments. It outlines the principal safety problems in these operations and provides the broad safety criteria which must be observed in the design, operation and maintenance of equipment and facilities for handling, transferring, and storing nuclear fuel and core components at nuclear power reactor sites

  10. Characteristics of fuel crud and its impact on storage, handling, and shipment of spent fuel

    International Nuclear Information System (INIS)

    Hazelton, R.F.

    1987-09-01

    Corrosion products, called ''crud,'' form on out-of-reactor surfaces of nuclear reactor systems and are transported by reactor coolant to the core, where they deposit on external fuel-rod cladding surfaces and are activated by nuclear reactions. After discharge of spent fuel from a reactor, spallation of radioactive crud from the fuel rods could impact wet or dry storage operations, handling (including rod consolidation), and shipping. It is the purpose of this report to review earlier (1970s) and more recent (1980s) literature relating to crud, its characteristics, and any impact it has had on actual operations. Crud characteristics vary from reactor type to reactor type, reactor to reactor, fuel assembly to fuel assembly in a reactor, circumferentially and axially in an assembly, and from cycle to cycle for a specific facility. To characterize crud of pressurized-water (PWRs) and boiling-water reactors (BWRs), published information was reviewed on appearance, chemical composition, areal density and thickness, structure, adhesive strength, particle size, and radioactivity. Information was also collected on experience with crud during spent fuel wet storage, rod consolidation, transportation, and dry storage. From experience with wet storage, rod consolidation, transportation, and dry storage, it appears crud spallation can be managed effectively, posing no significant radiological problems. 44 refs., 11 figs

  11. Development of nuclear fuel cycle remote handling technology

    International Nuclear Information System (INIS)

    Kim, K. H.; Park, B. S.; Kim, S. H.

    2010-04-01

    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

  12. Multi Canister Overpack (MCO) Handling Machine Independent Review of Seismic Structural Analysis

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-09-22

    The following separate reports and correspondence pertains to the independent review of the seismic analysis. The original analysis was performed by GEC-Alsthom Engineering Systems Limited (GEC-ESL) under subcontract to Foster-Wheeler Environmental Corporation (FWEC) who was the prime integration contractor to the Spent Nuclear Fuel Project for the Multi-Canister Overpack (MCO) Handling Machine (MHM). The original analysis was performed to the Design Basis Earthquake (DBE) response spectra using 5% damping as required in specification, HNF-S-0468 for the 90% Design Report in June 1997. The independent review was performed by Fluor-Daniel (Irvine) under a separate task from their scope as Architect-Engineer of the Canister Storage Building (CSB) in 1997. The comments were issued in April 1998. Later in 1997, the response spectra of the Canister Storage Building (CSB) was revised according to a new soil-structure interaction analysis and accordingly revised the response spectra for the MHM and utilized 7% damping in accordance with American Society of Mechanical Engineers (ASME) NOG-1, ''Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).'' The analysis was re-performed to check critical areas but because manufacturing was underway, designs were not altered unless necessary. FWEC responded to SNF Project correspondence on the review comments in two separate letters enclosed. The dispositions were reviewed and accepted. Attached are supplier source surveillance reports on the procedures and process by the engineering group performing the analysis and structural design. All calculation and analysis results are contained in the MHM Final Design Report which is part of the Vendor Information File 50100. Subsequent to the MHM supplier engineering analysis, there was a separate analyses for nuclear safety accident concerns that used the electronic input data files provided by FWEC/GEC-ESL and are contained in

  13. Spent Fuel Handling and Packaging Program: a survey of hot cell facilities

    International Nuclear Information System (INIS)

    Menon, M.N.

    1978-07-01

    Hot cell facilities in the United States were surveyed to determine their capabilities for conducting integral fuel assembly and individual fuel rod examinations that are required in support of the Spent Fuel Handling and Packaging Program. The ability to receive, handle, disassemble and reconstitute full-length light water reactor spent fuel assemblies, and the ability to conduct nondestructive and destructive examinations on full-length fuel rods were of particular interest. Three DOE-supported facilities and three commercial facilities were included in the survey. This report provides a summary of the findings

  14. Potential uses of remote handling and robotic techniques in the back end of the fuel cycle

    International Nuclear Information System (INIS)

    Reynolds, N.P.; Tabe, T.; Fenton, N.; Baumgartner, P.

    1984-01-01

    Atomic Energy of Canada Limited (AECL) is actively conducting research on used fuel immobilization, used fuel reprocessing, and nuclear fuel waste immobilization and disposal. This paper attempts to identify potential uses of robotics and remote handling techniques in these areas, where their adoption could lead to significant processing, economic and safety advantages

  15. Remote, under-sodium fuel handling experience at EBR-II

    International Nuclear Information System (INIS)

    King, R.W.; Planchon, H.P.

    1995-01-01

    The EBR-II is a pool-type design; the reactor fuel handling components and entire primary-sodium coolant system are submerged in the primary tank, which is 26 feet in diameter, 26 feet high, and contains 86,000 gallons of sodium. Since the reactor is submerged in sodium, fuel handling operations must be performed blind, making exact positioning and precision control of the fuel handling system components essential. EBR-II operated for 30 years, and the fuel handling system has performed approximately 25,000 fuel transfer operations in that time. Due to termination of the IFR program, EBR-II was shut down on September 30, 1994. In preparation for decommissioning, all fuel in the reactor will be transferred out of EBR-II to interim storage. This intensive fuel handling campaign will last approximately two years, and the number of transfers will be equivalent to the fuel handling done over about nine years of normal reactor operation. With this demand on the system, system reliability will be extremely important. Because of this increased demand, and considering that the system has been operating for about 32 years, system upgrades to increase reliability and efficiency are proceeding. Upgrades to the system to install new digital, solid state controls, and to take advantage of new visualization technology, are underway. Future reactor designs using liquid metal coolant will be able to incorporate imaging technology now being investigated, such as ultraviolet laser imaging and ultrasonic imaging

  16. Modern power station practice mechanical boilers, fuel-, and ash-handling plant

    CERN Document Server

    Sherry, A; Cruddace, AE

    2014-01-01

    Modern Power Station Practice, Second Edition, Volume 2: Mechanical (Boilers, Fuel-, and Ash-Handling Plant) focuses on the design, manufacture and operation of boiler units and fuel-and ash-handling plants.This book is organized into five main topics-furnace and combustion equipment, steam and water circuits, ancillary plant and fittings, dust extraction and draught plant, and fuel-and ash-handling plant.In these topics, this text specifically discusses the influence of nature of coal on choice of firing equipment; oil-burner arrangements, ignition and control; disposition of the heating surf

  17. Handling system for nuclear fuel cans to a fuel pellet feeder

    International Nuclear Information System (INIS)

    Vere, B.; Mathevon, P.

    1985-01-01

    The handling system comprises a first array of conveyors which takes a batch of casings from a delivery rack, alters the spacing between the casings, and delivers them to a vibrating table feeder, a second array of conveyors which readjusts the spacing between casing to its initial value and transfers the casings to a removal rack, and automatic and synchronized control means for ensuring the displacements of casings always in the same direction. The increase of spacing between casings can be used, before feeding, to allow them to be weighed one after the other, and after feeding, for cleaning the end part of fuel cans [fr

  18. Machine Vision Tests for Spent Fuel Scrap Characteristics

    International Nuclear Information System (INIS)

    BERGER, W.W.

    2000-01-01

    The purpose of this work is to perform a feasibility test of a Machine Vision system for potential use at the Hanford K basins during spent nuclear fuel (SNF) operations. This report documents the testing performed to establish functionality of the system including quantitative assessment of results. Fauske and Associates, Inc., which has been intimately involved in development of the SNF safety basis, has teamed with Agris-Schoen Vision Systems, experts in robotics, tele-robotics, and Machine Vision, for this work

  19. Fuel handling, reprocessing, and waste and related nuclear data aspects

    International Nuclear Information System (INIS)

    Kuesters, H.; Lalovic, M.; Wiese, H.W.

    1979-06-01

    The essential processes in the out-of-pile nuclear fuel cycle are described, i.e. mining and milling of uranium ores, enrichment, fuel fabrication, storage, transportation, reprocessing of irradiated fuel, waste treatment and waste disposal. The aspects of radiation (mainly gammas and neutrons) and of heat production, as well as special safety considerations are outlined with respect to their potential operational impacts and long-term hazards. In this context the importance of nuclear data for the out-of-pile fuel cycle is discussed. Special weight is given to the LWR fuel cycle including recycling; the differences of LMFBR high burn-up fuel with large PuO 2 content are described. The HTR fuel cycle is discussed briefly as well as some alternative fuel cycle concepts. (orig.) [de

  20. Micro manipulators to handle micro machines. ; Aiming at developing clever and deft robots. Micro machine wo handling surutameno micro manipulator. ; Kashikoku kiyo na robot kaihatsu wo mezashite

    Energy Technology Data Exchange (ETDEWEB)

    Fujie, M [Hitachi, Ltd., Tokyo (Japan)

    1991-11-15

    The current state of micro manipulators (MM) to control micro machines is described. No MM can be realized with the conventional robots of which finger positions and attitudes are controlled by information on angles of each articulate. To solve this problem, such a system using the conception called a master slave manipulator with complex structure is proposed, in which human ways of manipulation are coupled with master arms with simple structure and slave arms facilitating works via a computer. The MM requires a flexible articulate chaining mechanism when the considerations on smallness of the arm structure and chaining of the multiple actuators are taken. To control the relative positions and attitudes, development of the control algorithm is required, which can learn the command signals given to a large number of actuators to drive mechanisms with ultra high freedoms and the relationship among changes in the end effector movements, rotation and directions, and can perform works according to precise relative positioning of objects to be handled. 8 refs., 4 figs.

  1. Design support document for the K Basins Vertical Fuel Handling Tools

    International Nuclear Information System (INIS)

    Bridges, A.E.

    1995-01-01

    The purpose of this document is to provide the design support information for the Vertical Fuel Handling Tools, developed for the removal of N Reactor fuel elements from their storage canisters in the K Basins storage pool and insertion into the Single Fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N Reactor fuel elements is part of the overall characterization effort. These new hand tools are required since previous fuel movement has involved grasping the fuel in a horizontal position. These tools are required to lift an element vertically from the storage canister. Additionally, a Mark II storage canister Lip Seal Protector was designed and fabricated for use during fuel retrieval. This device was required to prevent damage to the canister lip should a fuel element accidentally be dropped during its retrieval, using the handling tools. Supporting documentation for this device is included in this document

  2. Fuel handling at Cernavoda 1 N.P.S. - commissioning and training philosophy

    Energy Technology Data Exchange (ETDEWEB)

    Standen, G W [AECL-Ansaldo Consortium, Cernavoda (Romania); Tiron, C; Marinescu, S [Regia Nationala de Electricitate (RENEL), Cernavoda (Romania); [Filiala Centrala Nuclearo Electrica (FCNE), Cernavoda (Romania)

    1997-12-31

    Efficient operation of a Candu nuclear power plant depends greatly on the reliable and safe operation of the fuel handling system. Successful commissioning of the system is obviously a key aspect of the reliability of the system and this coupled with a rigorous training programme for the fuel handling staff will ensure the system`s safe operation. This paper describes the philosophy used at Cernavoda 1 N.P.S. for the commissioning of the fuel handling systems and for the training of staff for operation and maintenance of these systems. The paper also reviews the commissioning programme, describing the milestones achieved and discussing some of the more interesting technical aspects which includes some unique Romanian input. In conclusion the paper looks at the organization of the mature fuel handling department from the operations, maintenance and technical support points of view and the long term plans for the future. (author). 1 fig.

  3. Fuel handling at Cernavoda 1 N.P.S. - commissioning and training philosophy

    International Nuclear Information System (INIS)

    Standen, G.W.; Tiron, C.; Marinescu, S.

    1996-01-01

    Efficient operation of a Candu nuclear power plant depends greatly on the reliable and safe operation of the fuel handling system. Successful commissioning of the system is obviously a key aspect of the reliability of the system and this coupled with a rigorous training programme for the fuel handling staff will ensure the system's safe operation. This paper describes the philosophy used at Cernavoda 1 N.P.S. for the commissioning of the fuel handling systems and for the training of staff for operation and maintenance of these systems. The paper also reviews the commissioning programme, describing the milestones achieved and discussing some of the more interesting technical aspects which includes some unique Romanian input. In conclusion the paper looks at the organization of the mature fuel handling department from the operations, maintenance and technical support points of view and the long term plans for the future. (author). 1 fig

  4. Fuel exchanging machine for a nuclear ship

    International Nuclear Information System (INIS)

    Hayashi, Tetsuji.

    1984-01-01

    Purpose: To prevent atmospheric contaminations upon fuel exchange thereby keep the environmental circumstance clean in the periphery of the nuclear ship. Constitution: A nuclear reactor container is disposed to the inside of a containing vessel in the ship body and a shutter is mounted to the upper opening of the ship body. Further, a landing container having a bottom opening equipped with shutter for alingning the upper opening equipped with shuuter of the ship is elevatably suspended to the trolley of a crane by way of a wire rope and a winch, and a fuel exchange cask is elevatably disposed to the inside of the landing container. Further, airs in the inside of the container is adapted to be discharged externally through a filter by means of a blower and the inside is kept at a negative pressure. Thus, since the containing vessel is covered with the landing container upon fuel exchanging operation, atmospheric contamination can be prevented sufficiently. (Sekiya, K.)

  5. Handling and carrying head for nuclear fuel assemblies and installation including this head

    International Nuclear Information System (INIS)

    Artaud, R.; Cransac, J.P.; Jogand, P.

    1986-01-01

    The present invention proposes a handling and carrying head ensuring efficiently the cooling of the nuclear fuel asemblies it transports so that any storage in liquid metal in a drum within or adjacent the reactor vessel is suppressed. The invention claims also a nuclear fuel handling installation including the head; it allows a longer time between loading and unloading campaigns and the space surrounding the reactor vessel keeps free without occupying a storage zone within the vessel [fr

  6. Evaluation of design and operation of fuel handling systems for 25 MW biomass fueled CFB power plants

    International Nuclear Information System (INIS)

    Precht, D.

    1991-01-01

    Two circulating fluidized bed, biomass fueled, 25MW power plants were placed into operation by Thermo Electron Energy Systems in California during late 1989. This paper discusses the initial fuel and system considerations, system design, actual operating fuel characterisitics, system operation during the first year and modifications. Biomass fuels handled by the system include urban/manufacturing wood wastes and agricultural wastes in the form of orchard prunings, vineyard prunings, pits, shells, rice hulls and straws. Equipment utilized in the fuel handling system are described and costs are evaluated. Lessons learned from the design and operational experience are offered for consideration on future biomass fueled installations where definition of fuel quality and type is subject to change

  7. 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)

  8. Analysis of tritium mission FMEF/FAA fuel handling accidents

    Energy Technology Data Exchange (ETDEWEB)

    Van Keuren, J.C.

    1997-11-18

    The Fuels Material Examination Facility/Fuel Assembly Area is proposed to be used for fabrication of mixed oxide fuel to support the Fast Flux Test Facility (FFTF) tritium/medical isotope mission. The plutonium isotope mix for the new mission is different than that analyzed in the FMEF safety analysis report. A reanalysis was performed of three representative accidents for the revised plutonium mix to determine the impact on the safety analysis. Current versions computer codes and meterology data files were used for the analysis. The revised accidents were a criticality, an explosion in a glovebox, and a tornado. The analysis concluded that risk guidelines were met with the revised plutonium mix.

  9. Extract of the report of the working party on the handling of irradiated fuel

    International Nuclear Information System (INIS)

    Berest, P.

    1983-01-01

    The French government has requested a working party with Prof. Neel in the chair to submit a report on the handling of irradiated fuel. This part of the report concerns the retreated fuels. It gives important elements for the debate and formulates recommendations for radioactive waste management [fr

  10. Proceedings of the 1. international conference on CANDU fuel handling systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    Besides information on fuel loading and handling systems for CANDU and PHWR reactors, the 25 papers in these proceedings also include some on dry storage, modification to fuel strings at Bruce A, and on the SLAR (spacer location and repositioning) system for finding and moving garter springs. The individual papers have been abstracted separately.

  11. Proceedings of the 1. international conference on CANDU fuel handling systems

    International Nuclear Information System (INIS)

    1996-01-01

    Besides information on fuel loading and handling systems for CANDU and PHWR reactors, the 25 papers in these proceedings also include some on dry storage, modification to fuel strings at Bruce A, and on the SLAR (spacer location and repositioning) system for finding and moving garter springs. The individual papers have been abstracted separately

  12. 29 CFR 1917.156 - Fuel handling and storage.

    Science.gov (United States)

    2010-07-01

    ...) Liquid fuel dispensing devices, such as pumps, shall be mounted either on a concrete island or be...) Containers shall be examined before recharging and again before reuse for the following: (A) Dents, scrapes...

  13. FEMA and RAM Analysis for the Multi Canister Overpack (MCO) Handling Machine

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    The Failure Modes and Effects Analysis and the Reliability, Availability, and Maintainability Analysis performed for the Multi-Canister Overpack Handling Machine (MHM) has shown that the current design provides for a safe system, but the reliability of the system (primarily due to the complexity of the interlocks and permissive controls) is relatively low. No specific failure modes were identified where significant consequences to the public occurred, or where significant impact to nearby workers should be expected. The overall reliability calculation for the MHM shows a 98.1 percent probability of operating for eight hours without failure, and an availability of the MHM of 90 percent. The majority of the reliability issues are found in the interlocks and controls. The availability of appropriate spare parts and maintenance personnel, coupled with well written operating procedures, will play a more important role in successful mission completion for the MHM than other less complicated systems

  14. Remote waste handling at the Hot Fuel Examination Facility

    International Nuclear Information System (INIS)

    Vaughn, M.E.

    1982-01-01

    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

  15. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  16. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  17. Study over problems related to fuel and ash handling systems; Probleminventering braensle- och askhantering

    Energy Technology Data Exchange (ETDEWEB)

    Njurell, Rolf; Wikman, Karin [AaF-Energi och Miljoe AB, Stockhom (Sweden)

    2003-10-01

    There have been a lot of problems related to fuel and ash handling systems since the combustion of different types of biofuels started in the 70s. Many measures have been taken to solve some of the problems, but others have become part of the daily work. The purpose of this study has been to do a compilation of the fuel and ash handling problems that exist at different types of heat and power plants. The study over problems related to fuel and ash handling systems has been carried out through a questionnaire via the Internet. Directors at about 150 energy production plants were contacted by phone or e-mail in the beginning of the project and asked to participate in the study. 72 of these plants accepted to fill in the questionnaire. After several reminders by e-mails and phone calls there were in the end 32 plants that completed the form. Together they reported about 25 problems related to fuel handling and 27 problems related to ash handling. In general each of the plants reported one problem of each kind. Even if the material from the questionnaire is not enough to make statistical analysis a few conclusions can be made about the most common problems, the cause of the problems and where they appear. Fuel handling problems that occur at several plants are stoppage in the conveying equipment, bridging in the boiler silo or the tipping bunker and problems with the sieve for separation. The distribution of the fuel handling problems is almost equal for all equipment parts (receiving, separation, transport etc.). For the ash handling systems problems with transport of dry bottom ash dominate, followed by and the moistening of fly ash and transport of wet bottom ash. Most of the problems related to fuel handling are caused by the fuel quality. For example several plants have reported that bark is a fuel that is hard to handle. Nevertheless the quality for a specific fuel is not always bad when it is delivered to the plant but the fuel quality might change during

  18. Remote Handling Devices for Disposition of Enriched Uranium Reactor Fuel Using Melt-Dilute Process

    International Nuclear Information System (INIS)

    Heckendorn, F.M.

    2001-01-01

    Remote handling equipment is required to achieve the processing of highly radioactive, post reactor, fuel for the melt-dilute process, which will convert high enrichment uranium fuel elements into lower enrichment forms for subsequent disposal. The melt-dilute process combines highly radioactive enriched uranium fuel elements with deleted uranium and aluminum for inductive melting and inductive stirring steps that produce a stable aluminum/uranium ingot of low enrichment

  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. Work plan for development of K-Basin fuel handling tools

    International Nuclear Information System (INIS)

    Bridges, A.E.

    1994-01-01

    The purpose of this document is to provide the engineering work plan for the development of handling tools for the removal of N-Reactor fuel elements from their storage canisters in the K-Basins storage pool and insertion into the Single Fuel Element Cans for subsequent shipment to a Hot Cell for examination. Examination of these N-Reactor fuel elements is part of the overall characterization effort. New hand tools are required since previous fuel movement has involved grasping the fuel in a horizontal position. These tools are required to lift an element from the storage canister

  1. Challenges and innovative technologies on fuel handling systems for future sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements. (author)

  2. Device for handling fuel assemblies within a nuclear reactor core

    International Nuclear Information System (INIS)

    Dupuy, G.

    1975-01-01

    A device is described which comprises two arms having synchronized movements, the tubular portions of the two arms being rigidly fixed to each other by means of a sliding connection and capable of being endowed by means of a differential drive system with movements for producing the same effort but applied in the opposite direction in order that the lateral reaction on the grab of the first arm on a fuel assembly should be equal and opposite to the lateral reaction of the cylindrical guide tube of the second arm on the same fuel assembly

  3. Computer control of fuel handling activities at FFTF

    International Nuclear Information System (INIS)

    Romrell, D.M.

    1985-03-01

    The Fast Flux Test Facility near Richland, Washington, utilizes computer control for reactor refueling and other related core component handling and processing tasks. The computer controlled tasks described in this paper include core component transfers within the reactor vessel, core component transfers into and out of the reactor vessel, remote duct measurements of irradiated core components, remote duct cutting, and finally, transferring irradiated components out of the reactor containment building for off-site shipments or to long term storage. 3 refs., 16 figs

  4. Handling system for nuclear reactor fuel and reflector elements

    International Nuclear Information System (INIS)

    Hawke, B.C.; Goldman, L.A.

    1980-01-01

    A system for canning, inspecting and transferring to a storage area fuel and reflector elements from a nuclear reactor is described. The canning mechanism operates in a sealed gaseous environment and visual and mechanical inspection of the elements is possible by an operator from a remote shielded area. (UK)

  5. Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 2. Appendices

    International Nuclear Information System (INIS)

    1979-08-01

    This volume contains the following appendices: LWR fuel cycle, handling and storage of spent fuel, termination case considerations (use of coal-fired power plants to replace nuclear plants), increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data, characteristics of nuclear fuel, away-from-reactor storage concept, spent fuel storage requirements for higher projected nuclear generating capacity, and physical protection requirements and hypothetical sabotage events in a spent fuel storage facility

  6. Hot fuel examination facility element spacer wire-wrap machine

    International Nuclear Information System (INIS)

    Tobias, D.A.; Sherman, E.K.

    1989-01-01

    Nondestructive examinations of irradiated experimental fuel elements conducted in the Argonne National Laboratory Hot Fuel Examination Facility/North (HFEF/N) at the Idaho National Engineering Laboratory include laser and contact profilometry (element diameter measurements), electrical eddy-current testing for cladding and thermal bond defects, bow and length measurements, neutron radiography, gamma scanning, remote visual exam, and photography. Profilometry was previously restricted to spiral profilometry of the element to prevent interference with the element spacer wire wrapped in a helix about the Experimental Breeder Reactor II (EBR-II)-type fuel element from end to end. By removing the spacer wire prior to conducting profilometry examination, axial profilometry techniques may be used, which are considerably faster than spiral techniques and often result in data acquisition more important to experiment sponsors. Because the element must often be reinserted into the nuclear reactor (EBR-II) for additional irradiation, however, the spacer wire must be reinstalled on the highly irradiated fuel element by remote means after profilometry of the wireless elements. The element spacer wire-wrap machine developed at HFEF is capable of helically wrapping fuel elements with diameters up to 1.68 cm (0.660 in.) and 2.44-m (96-in.) lengths. The machine can accommodate almost any desired wire pitch length by simply inserting a new wrapper gear module

  7. Experience with fuel damage caused by abnormal conditions in handling and transporting operations

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1983-01-01

    Pacific Northwest Laboratory (PNL) conducted a study to determine the expected condition of spent USA light-water reactor (LWR) fuel upon arrival at interim storage or fuel reprocessing facilities or, if fuel is declared a waste, at disposal facilities. Initial findings were described in an earlier PNL paper at PATRAM '80 and in a report. Updated findings are described in this paper, which includes an evaluation of information obtained from the literature and a compilation of cases of known or suspected damage to fuel as a result of handling and/or transporting operations. To date, PNL has evaluated 123 actual cases (98 USA and 25 non-USA). Irradiated fuel was involved in all but 10 of the cases. From this study, it is calculated that the frequency of unusual occurrences involving fuel damage from handling and transporting operations has been low. The damage that did occur was generally minor. The current base of experience with fuel handling and transporting operations indicates that nearly all of these unusual occurrences had only a minor or negligible effect on spent fuel storage facility operations

  8. Handling apparatus for fuel assemblies in a core

    International Nuclear Information System (INIS)

    Hatakenaka, Hideo.

    1975-01-01

    Object: To prevent an occurrence of a cloud as well as trouble in outflow of cooling water at the time of failure, in a window through which the operation of a collet installing and removing mechanism is monitored. Structure: A monitoring window comprises a pair of transparent window panes between which is interposed a non-compressive transparent fluid. With this construction, when the collet installing and removing mechanism within a container is operated while illuminating it by light means and monitoring it by a television camera to connect a fuel assembly with a shielding plug, and even if one transparent window pane should be failed as a result of trouble, the other transparent window pane prevents outflow of cooling water within a fuel transferring transfer port, and at the same time, the scattering force of fragments of failed transparent window pane is attenuated by the non-compressive transparent body within the monitoring window chamber. (Hanada, M.)

  9. Simulator for candu600 fuel handling system. environmental implications

    International Nuclear Information System (INIS)

    Vulpe, S.; Valeca, S.; Predescu, D.

    2016-01-01

    Personnel training are a main topic in the security and reliability of several industrial processes. The simulator is a physical device that reproduces real operation of a device used in a production process technology. Typically, a simulator is intended to train the operators to work properly with the real device in the production process, but simulators can be involved frequently in the research and evaluation of performance of human operators. Process simulation has a significant role in the training of operators of nuclear plants. To ensure the safe operation, protection of workers and the environment, of any nuclear power plant, the training of its operators in all operating modes of the plant is essential. A trained operator who can handle any emergency in a controlled manner, without panic, protecting equipment and personnel is an asset for a nuclear power plant. (authors)

  10. Spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    International Nuclear Information System (INIS)

    Daling, P.M.; Konzek, G.J.; Lezberg, A.J.; Votaw, E.F.; Collingham, M.I.

    1985-04-01

    This report describes an evaluation of the cask handling capabilities of those reactors which are operating or under construction. A computerized data base that includes cask handling information was developed with information from the literature and utility-supplied data. The capability of each plant to receive and handle existing spent fuel shipping casks was then evaluated. Modal fractions were then calculated based on the results of these evaluations and the quantities of spent fuel projected to be generated by commercial nuclear power plants through 1998. The results indicated that all plants are capable of receiving and handling truck casks. Up to 118 out of 130 reactors (91%) could potentially handle the larger and heavier rail casks if the maximum capability of each facility is utilized. Design and analysis efforts and physical modifications to some plants would be needed to achieve this high rail percentage. These modifications would be needed to satisfy regulatory requirements, increase lifting capabilities, develop rail access, or improve other deficiencies. The remaining 12 reactors were determined to be capable of handling only the smaller truck casks. The percentage of plants that could receive and handle rail casks in the near-term would be reduced to 64%. The primary reason for a plant to be judged incapable of handling rail casks in the near-term was a lack of rail access. The remaining 36% of the plants would be limited to truck shipments. The modal fraction calculations indicated that up to 93% of the spent fuel accumulated by 1998 could be received at federal storage or disposal facilities via rail (based on each plant's maximum capabilities). If the near-term cask handling capabilities are considered, the rail percentage is reduced to 62%

  11. Safety of handling, storing and transportation of spent nuclear fuel and vitrified high-level wastes

    International Nuclear Information System (INIS)

    Ericsson, A.M.

    1977-11-01

    The safety of handling and transportation of spent fuel and vitrified high-level waste has been studied. Only the operations which are performed in Sweden are included. That is: - Transportation of spent fuel from the reactors to an independant spent fuel storage installation (ISFSI). - Temporary storage of spent fuel in the ISFSI. - Transportation of the spent fuel from the ISFSI to a foreign reprocessing plant. - Transportation of vitrified high-level waste to an interim storage facility. - Interim storage of vitrified high-level waste. - Handling of the vitrified high-level waste in a repository for ultimate disposal. For each stage in the handling sequence above the following items are given: - A brief technical description. - A description of precautionary measures considered in the design. - An analysis of the discharges of radioactive materials to the environment in normal operation. - An analysis of the discharges of radioactive materials due to postulated accidents. The dose to the public has been roughly and conservatively estimated for both normal and accident conditions. The expected rate of occurence are given for the accidents. The results show that above described handling sequence gives only a minor risk contribution to the public

  12. MCO Engineering Test Report Fuel Basket Handling Grapple Acceptance Test

    International Nuclear Information System (INIS)

    CHENAULT, D.M.

    2000-01-01

    Acceptance testing of the production SNF Fuel Basket lift grapples to the required 150 percent maximum lift load is documented herein. The report shows the results affirming the proof test passage. The primary objective of this test was to confirm the load rating of the grapple per applicable requirements of ANSI 14 6 American National Standard For Radioactive Materials Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500kg) or More. The above Standard requires a load test of 150% of the design load which must be held for a minimum of 10 minutes followed by a Liquid Penetrant or Magnetic Particle examination of critical areas and welds in accordance with the ANSI/ASME Boiler and Pressure Vessel Code 1989 Section 111 Division 1 section NF 5350

  13. Operating experiences in fuel handling system at KGS

    International Nuclear Information System (INIS)

    Reddy, G.P.; Nagabhushanam

    2006-01-01

    Refuelling operations were started at KGS in August, 2000. Rich and varied experience was gained during this period through internal discussion/Quality circles/Procedural reviews and analysis of various incidents that have taken place in KGS and other units of NPCIL Some of the unique jobs carried out at KGS include-Development of tools for in-situ replacement of FM front end cover in FM service area (which was done for the first time in NPCIL history), Modification of FM magazine rear end plate mounting screws to avoid the possibility of magazine rotation stalling, The incident of Stalling of B-Ram during installation of upstream shield plug in KGS - 1 has brought out many weakness that were existing in the system in a dormant manner. Review of maintenance procedures was carried out and a special underwater operated sensor was developed and installed in Transfer Magazine to sense the presence and proper positioning of fuel bundles in the Transfer magazine tube during fuel loading operation. Numerous modifications were carried out in the system to increase equipment reliability, ease of operation and maintenance, to reduce man-rem consumption. Most notable among these modifications include -zig saw panel modification, EFCV O-ring modification, Ram BF switch modification, provision for increase in SFSB level provision, snout clamp oil circuit modification, ball valve actuator modification, installation of additional switch for sensing STS carriage UP position etc, This paper focuses on the challenges tackled in achieving near perfect performance, innovations and improvements carried out in the system to strive for this goal and development of procedures for reducing man-rem consumption and life extension of critical components. (author)

  14. Preparation for commissioning of nuclear plant with reference to British Nuclear Fuels Plc fuel handling plant project

    International Nuclear Information System (INIS)

    Bamber, D.R.

    1987-01-01

    The new Fuel Handling Plant at British Nuclear Fuels Sellafield is part of a Pound 550M complex which provides facilities for the receipt, storage and mechanical preparation of both Magnox and A.G.R. fuel. The plant is very large and complex with considerable use of computer based process control systems, providing for physical and nuclear safety. The preparation of such plant for active commissioning necessitates a great many physical checks and technical evaluations in support of its safety case. This paper describes arrangements for plant commissioning checks, against the regulatory framework and explains the physical preparations necessary for their timely accomplishment. (author)

  15. A study of the effectiveness of hand protection when handling UO2 fuel pellets

    International Nuclear Information System (INIS)

    Washington, R.R.; Sullivan, D.F.

    1981-01-01

    Simple tests were performed to estimate the effectiveness of various forms of hand protection in reducing skin doses when handling UO 2 fuel pellets. Household rubber gloves (rubberized cotton) appeared to be the most effective of the varieties tested. Nylon gloves and latex finger cots were least effective. (author)

  16. 18 CFR 1304.405 - Fuel storage tanks and handling facilities.

    Science.gov (United States)

    2010-04-01

    ... used to contain a regulated substance (such as a petroleum product) and has 10 percent or more of its... or remedy pollution or violations of law, including removal of the UST system, with costs charged to... flammable and combustible liquids storage tanks at marine service stations. (d) Fuel handling on private...

  17. The FMEA Analysis for Fuel Handling System at Cernavoda Unit 2

    International Nuclear Information System (INIS)

    Park, Jin Hee; Kim, Tae Woon; Rhee, Bo Wook; Yoon, Chul; Kim, Hyeong Tae; Cho, In Gil; Kim, Seong Ho

    2006-01-01

    A Nuclear Safety Evaluation was performed by an independent assessor at the request of the regulatory authority CNCAN (Comisia Nationala pentru Controlul Activitatilor Nucleare. National Committee for Nuclear Activities Control in Romania) to provide an independent overview of all the nuclear safety aspects of Cernavoda Unit 2 under construction and an expert opinion whether the completed Cernavoda Unit-2 Nuclear Power Plant would satisfy current Western European nuclear safety objectives and practices. A report was produced (Cernavoda 2 Nuclear Safety Expert Project, 'Task 10 . Safety Evaluation Report', A.F.Parsons, NNC Limited, December 2001) and contains recommendations either mandatory or advisory. The FMEA study, one of the mandatory recommendations, is performing now for fuel handling system and radioactive waste handling system for Cernavoda unit 2 in Romania sponsored by KHNP. In this paper, only the FMEA study for fuel handling system is presented

  18. 18 years experience on UF{sub 6} handling at Japanese nuclear fuel manufacturer

    Energy Technology Data Exchange (ETDEWEB)

    Fujinaga, H.; Yamazaki, N.; Takebe, N. [Japan Nucelar Fuel Conversion Co., Ltd., Ibaraki (Japan)

    1991-12-31

    In the spring of 1991, a leading nuclear fuel manufacturing company in Japan, celebrated its 18th anniversary. Since 1973, the company has produced over 5000 metric ton of ceramic grade UO{sub 2} powder to supply to Japanese fabricators, without major accident/incident and especially with a successful safety record on UF{sub 6} handling. The company`s 18 years experience on nuclear fuel manufacturing reveals that key factors for the safe handling of UF{sub 6} are (1) installing adequate facilities, equipped with safety devices, (2) providing UF{sub 6} handling manuals and executing them strictly, and (3) repeating on and off the job training for operators. In this paper, equipment and the operation mode for UF{sub 6} processing at their facility are discussed.

  19. Towards a better mastery of risks in the handling of nuclear fuel: the contributions of ergonomics

    International Nuclear Information System (INIS)

    Samson, L.

    1999-01-01

    Nuclear fuel is handled under water in the reactor pool using procedures that have yet to be automated. The knowledge and skill of the operators is therefore of prime importance. Ergonomic consultants have prepared a report on the problems facing the operators when handling nuclear fuel? These problems have been addressed by the installation of a new system to detect and prevent incorrect operator commands and to provide software assistance in planning movements together with diagnostic functions. The new system has resulted in considerable time savings and a reduction in the risk of error. However, it has been necessary to modify the control software in the light of the handling strategies traditionally used by the operators. (author)

  20. Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

    Energy Technology Data Exchange (ETDEWEB)

    Madhuresh, R; Nagarajan, R; Jit, I; Sanatkumar, A [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like `stay-put`, `gravity- fill`, `D{sub 2}0- steaming` etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs.

  1. Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

    International Nuclear Information System (INIS)

    Madhuresh, R.; Nagarajan, R.; Jit, I.; Sanatkumar, A.

    1996-01-01

    Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like 'stay-put', 'gravity- fill', 'D 2 0- steaming' etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs

  2. Development of CANDU prototype fuel handling simulator - concept and some simulation results with physical network modeling approach

    Energy Technology Data Exchange (ETDEWEB)

    Xu, X.P. [Candu Energy Inc, Mississauga, Ontario (Canada)

    2012-07-01

    This paper reviewed the need for a fuel handling(FH) simulator in training operators and maintenance personnel, in FH design enhancement based on operating experience (OPEX), and the potential application of Virtual Reality (VR) based simulation technology. Modeling and simulation of the fuelling machine (FM) magazine drive plant (one of the CANDU FH sub-systems) was described. The work established the feasibility of modeling and simulating a physical FH drive system using the physical network approach and computer software tools. The concept and approach can be applied similarly to create the other FH subsystem plant models, which are expected to be integrated with control modules to develop a master FH control model and further to create a virtual FH system. (author)

  3. Development of CANDU prototype fuel handling simulator - concept and some simulation results with physical network modeling approach

    International Nuclear Information System (INIS)

    Xu, X.P.

    2012-01-01

    This paper reviewed the need for a fuel handling(FH) simulator in training operators and maintenance personnel, in FH design enhancement based on operating experience (OPEX), and the potential application of Virtual Reality (VR) based simulation technology. Modeling and simulation of the fuelling machine (FM) magazine drive plant (one of the CANDU FH sub-systems) was described. The work established the feasibility of modeling and simulating a physical FH drive system using the physical network approach and computer software tools. The concept and approach can be applied similarly to create the other FH subsystem plant models, which are expected to be integrated with control modules to develop a master FH control model and further to create a virtual FH system. (author)

  4. Management and Handling of Rejected Fuel of MTR Type and Process Effluents Contained Uranium at FEPI

    International Nuclear Information System (INIS)

    Ghaib Widodo; Bambang Herutomo

    2007-01-01

    Research Reactor Fuel Element Production Installation (FEPI) - Serpong has performed management and handling of all kinds of rejected fuel material during production (solids, liquids, and gases) and process effluents contained uranium. The methods that has been implemented are precipitation, absorption, evaporation, electrolysis, and electrodialysis. By these methods will finally be obtained forms of product which can be used directly as fuel material feed and solid/liquid radioactive waste that fulfil the requirements (uranium contents < 50 ppm) to be send to Radioactive Waste Management Installation. (author)

  5. Study on new-type fuel-related assembly handling tools for PWR NPP

    International Nuclear Information System (INIS)

    Fan Xiumei

    2013-01-01

    This article describes the design and study on a set of new-type fuel-related assembly snatching tools used for PWR NPP. The purpose is mainly to enhance the tool safety, reliability and convenientness by improvement of the mechanism and structure of the tool for snatching preciseness and avoiding from falling and abrasion of fuel-related assemblies for any condition. The new-type fuel-related assembly handling tools are compared with similar equipment in worldwide in terms of function, main technical characteristic, and safety and protection, some of them are better than the similar equipment in that they have reliable loading and unloading and conveying capabilities. (author)

  6. Multi Canister Overpack (MCO) Handling Machine Trolley Seismic Uplift Constraint Design Loads

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    The MCO Handling Machine (MHM) trolley moves along the top of the MHM bridge girders on east-west oriented rails. To prevent trolley wheel uplift during a seismic event, passive uplift constraints are provided as shown in Figure 1-1. North-south trolley wheel movement is prevented by flanges on the trolley wheels. When the MHM is positioned over a Multi-Canister Overpack (MCO) storage tube, east-west seismic restraints are activated to prevent trolley movement during MCO handling. The active seismic constraints consist of a plunger, which is inserted into slots positioned along the tracks as shown in Figure 1-1. When the MHM trolley is moving between storage tube positions, the active seismic restraints are not engaged. The MHM has been designed and analyzed in accordance with ASME NOG-1-1995. The ALSTHOM seismic analysis (Reference 3) reported seismic uplift restraint loading and EDERER performed corresponding structural calculations. The ALSTHOM and EDERER calculations were performed with the east-west seismic restraints activated and the uplift restraints experiencing only vertical loading. In support of development of the CSB Safety Analysis Report (SAR), an evaluation of the MHM seismic response was requested for the case where the east-west trolley restraints are not engaged. For this case, the associated trolley movements would result in east-west lateral loads on the uplift constraints due to friction, as shown in Figure 1-2. During preliminary evaluations, questions were raised as to whether the EDERER calculations considered the latest ALSTHOM seismic analysis loads (See NCR No. 00-SNFP-0008, Reference 5). Further evaluation led to the conclusion that the EDERER calculations used appropriate vertical loading, but the uplift restraints would need to be re-analyzed and modified to account for lateral loading. The disposition of NCR 00-SNFP-0008 will track the redesign and modification effort. The purpose of this calculation is to establish bounding seismic

  7. Evaluation of the Total Cost of Ownership of Fuel Cell-Powered Material Handling Equipment

    Energy Technology Data Exchange (ETDEWEB)

    Ramsden, T.

    2013-04-01

    This report discusses an analysis of the total cost of ownership of fuel cell-powered and traditional battery-powered material handling equipment (MHE, or more typically 'forklifts'). A number of fuel cell MHE deployments have received funding support from the federal government. Using data from these government co-funded deployments, DOE's National Renewable Energy Laboratory (NREL) has been evaluating the performance of fuel cells in material handling applications. NREL has assessed the total cost of ownership of fuel cell MHE and compared it to the cost of ownership of traditional battery-powered MHE. As part of its cost of ownership assessment, NREL looked at a range of costs associated with MHE operation, including the capital costs of battery and fuel cell systems, the cost of supporting infrastructure, maintenance costs, warehouse space costs, and labor costs. Considering all these costs, NREL found that fuel cell MHE can have a lower overall cost of ownership than comparable battery-powered MHE.

  8. Spent-fuel shipping and cask-handling studies in wet and dry environments. Studies and research concerning BNFP

    International Nuclear Information System (INIS)

    McCreery, P.N.

    1982-09-01

    A demonstration cask system has been constructed specifically to be used in examining unconventional techniques in handling spent fuel and fuel-hauling casks. This report demonstrates, through a series of photographs, some of these techniques and discusses others. It includes wet and dry operations, loading and unloading horizontally and vertically, mobile on-site carriers that can eliminate the need for some cranes and, in general, many of the operational options that are open in the design of future fuel handling systems

  9. Remote handling equipment for laboratory research of fuel reprocessing in Nuclear Research Institute at Rez

    International Nuclear Information System (INIS)

    Fidler, J.; Novy, P.; Kyrs, M.

    1985-04-01

    Laboratory installations were developed for two nuclear fuel reprocessing methods, viz., the solvent extraction process and the fluoride volatility process. The apparatus for solvent extraction reprocessing consists of a pneumatically driven rod-chopper, a dissolver, mixer-settler extractors, an automatic fire extinguishing device and other components and it was tested using irradiated uranium. The technological line for the fluoride volatility process consists of a fluorimater, condensers, sorption columns with NaF pellets and a distillation column for the separation of volatile fluorides from UF 6 . The line has not yet been tested using irradiated fuel. Some features of the remote handling equipment of both installations are briefly described. (author)

  10. CRBRP design and test results for fuel handling systems, plugs, and seals

    International Nuclear Information System (INIS)

    Berg, G.E.

    1977-01-01

    The fuel handling system and reactor rotating plugs for the Clinch River Breeder Reactor Plant (CRBRP) are based primarily on existing technology and, in many respects, follow the concept developed for the Fast Flux Test Facility (FFTF). The equipment and the development programs initiated to verify its performance are described. Test results obtained from the development program, and the extent to which these results verified original design selections, or suggested potential improvements, are discussed

  11. Handling and disposal of SP-100 ground test nuclear fuel and equipment

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Hodgson, R.D.

    1990-05-01

    The post SP-100 reactor testing period will focus on defueling the reactor, packaging the various radioactive waste forms, and shipping this material to the appropriate locations. Remote-handling techniques will be developed to defuel the reactor. Packaging the spent fuel and activated reactor components is a challenge in itself. This paper presents an overview of the strategy, methods, and equipment that will be used during the closeout phase of nuclear testing

  12. Handling and disposal of SP-100 ground test nuclear fuel and equipment

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Hodgson, R.D.

    1991-01-01

    The post SP-100 reactor testing period will focus on defueling the reactor, packaging the various radiactive waste forms, and shipping this material to the appropriate locations. Remote-handling techniques will be developed to defuel the reactor. Packaging the spent fuel and activated reactor components is a challenge in itself. This paper presents an overview of the strategy, methods, and equipment that will be used during the closeout phase of nuclear testing

  13. EBR-II argon cooling system restricted fuel handling I and C upgrade

    International Nuclear Information System (INIS)

    Start, S.E.; Carlson, R.B.; Gehrman, R.L.

    1995-01-01

    The instrumentation and control of the Argon Cooling System (ACS) restricted fuel handling control system at Experimental Breeder Reactor II (EBR-II) is being upgraded from a system comprised of many discrete components and controllers to a computerized system with a graphical user interface (GUI). This paper describes the aspects of the upgrade including reasons for the upgrade, the old control system, upgrade goals, design decisions, philosophies and rationale, and the new control system hardware and software

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    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.

  15. Examination on the safety of handling the fuel elements in the nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    1977-01-01

    This is the report of the Examination Committee on Total Inspection and Repair Technologies for Mutsu to the Director of Science and Technology Agency and the Minister of Transport dated July 29, 1977. The committee concluded before that the total inspection on safety and the repair of shielding can be carried out as the fuel elements are loaded, and the safety can be secured sufficiently. It was decided at the meeting of ministers concerned with Mutsu on May 17 that the safety concerning handling the fuel elements of Mutsu should be examined by the committee. Under the premise that the fuel elements are loaded again and used after the total inspection on safety and the repair of shielding, the committee examined the methods and the basic concept of safety about the taking-out, transport and preservation of the fuel elements, and the conclusions obtained are reported. The contents of the examination are the outline of the fuel elements, the present condition of the fuel elements, the safety concerning taking-out, transport and preservation of the fuel elements, and the other measures required for securing safety. The committee thinks that the safety can be secured sufficiently if the works are carried out carefully. (Kako, I.)

  16. Comparative economics for DUCRETE spent fuel storage cask handling, transportation, and capital requirements

    International Nuclear Information System (INIS)

    Powell, F.P.

    1995-04-01

    This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask's primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives

  17. Simulation of Multi Canister Overpack (MCO) Handling Machine Impact with Cask and MCO During Insertion into the Transfer Pit (FDT-137)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-04-13

    The K-Basin Cask and Transportation System will be used for safely packaging and transporting approximately 2,100 metric tons of unprocessed, spent nuclear fuel from the 105 K East and K West Basins to the 200 E Area Canister Storage Building (CSB). Portions of the system will also be used for drying the spent fuel under cold vacuum conditions prior to placement in interim storage. The spent nuclear fuel is currently stored underwater in the two K-Basins. The K-Basins loadout pit is the area selected for loading spent nuclear fuel into the Multi-Canister Overpack (MCO) which in turn is located within the transportation cask. This Cask/MCO unit is secured.in the pit with a pail load out structure whose primary function is lo suspend and support the Cask/MCO unit at.the desired elevations and to protect the unit from the contaminated K-Basin water. The fuel elements will be placed in special baskets and stacked in the MCO that have been previously placed in the cask. The casks will be removed from the K Basin load out areas and taken to the cold vacuum drying station. Then the cask will be prepared for transportation to the CSB. The shipments will occur exclusively on the Hanford Site between K-Basins and the CSB. Travel will be by road with one cask per trailer. At the CSB receiving area the cask will be removed from the trailer. A gantry crane will then move the cask over to the transfer pit and load the cask into the transfer pit. From the transfer pit the MCO will be removed from the cask by the MCO Handling Machine (MHM). The MHM will move the MCO from the transfer pit to a canister storage tube in the CSB. MCOs will be piled two high in each canister Storage tube.

  18. Simulation of Multi Canister Overpack (MCO) Handling Machine Impact with Cask and MCO During Insertion into the Transfer Pit (FDT-137)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2000-01-01

    The K-Basin Cask and Transportation System will be used for safely packaging and transporting approximately 2,100 metric tons of unprocessed, spent nuclear fuel from the 105 K East and K West Basins to the 200 E Area Canister Storage Building (CSB). Portions of the system will also be used for drying the spent fuel under cold vacuum conditions prior to placement in interim storage. The spent nuclear fuel is currently stored underwater in the two K-Basins. The K-Basins loadout pit is the area selected for loading spent nuclear fuel into the Multi-Canister Overpack (MCO) which in turn is located within the transportation cask. This Cask/MCO unit is secured.in the pit with a pail load out structure whose primary function is lo suspend and support the Cask/MCO unit at.the desired elevations and to protect the unit from the contaminated K-Basin water. The fuel elements will be placed in special baskets and stacked in the MCO that have been previously placed in the cask. The casks will be removed from the K Basin load out areas and taken to the cold vacuum drying station. Then the cask will be prepared for transportation to the CSB. The shipments will occur exclusively on the Hanford Site between K-Basins and the CSB. Travel will be by road with one cask per trailer. At the CSB receiving area the cask will be removed from the trailer. A gantry crane will then move the cask over to the transfer pit and load the cask into the transfer pit. From the transfer pit the MCO will be removed from the cask by the MCO Handling Machine (MHM). The MHM will move the MCO from the transfer pit to a canister storage tube in the CSB. MCOs will be piled two high in each canister Storage tube

  19. Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

    International Nuclear Information System (INIS)

    Diggs, J.M.

    1985-01-01

    This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables

  20. Development of remote handling technology for nuclear fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi; Sakai, Akira; Miura, Noriaki; Kozaka, Tetsuo; Hamada, Takashi

    2015-01-01

    Remote handling technology has been systematically developed for nuclear fuel cycle facilities in Japan since 1970s, primarily in parallel with the development of reprocessing and HLLW (High Level Liquid Waste) vitrification process. In case of reprocessing and vitrification process to handle highly radioactive and hazardous materials, the most of components are installed in the radiation shielded hot cells and operators are not allowed to enter the work area in the cells for operation and maintenance. Therefore, a completely remote handling system is adopted for the cells to reduce radiation doses of operators and increase the availability of the facility. The hot cells are generally designed considering the scale of components (laboratory, demonstration, or full-scale), the function of the systems (chemical process, material handling, dismantling, decontamination, or chemical analysis), and the environmental conditions (radiation dose rate, airborne concentration, surface contamination, or fume/mist/dust). Throughout our domestic development work for remote handling technology, the concept of the large scale integrated cell has been adopted rather than a number of small scale separated cells, for the reasons to reduce the total installation space and the number of remote handling equipment required for the each cell as much as possible. In our domestic remote maintenance design, several new concepts have been developed, tested, and demonstrated in the Tokai Virtrification Facility (TVF) and the Rokkasho HLLW Vitrification and Storage Facility (K-facility). Layout in the hot cells, the performance of remote handling equipment, and the structure of the in-cell components are important factors for remote maintenance design. In case of TVF (hot tests started in 1995), piping and vessels are prefabricated in the rack modules and installed in two lines on both sides of the cell. These modules are designed to be remotely replaced in the whole rack. Two overhead cranes

  1. Handling of spent nuclear fuel and final storage of nitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    The following stages of handling and transport of the fuel on its way to final storage are dealt with in the report. 1) The spent nuclear fuel is stored at the power station or in the central fuel storage facility awaiting reprocessing. 2) The fuel is reprocessed, i.e. uranium, plutonium and waste are separated from each other. Reprocessing does not take place in Sweden. The highlevel waste is vitrified and can be sent back to Sweden in the 1990s. 3) Vitrified waste is stored for about 30 years awaiting deposition in the final repository. 4) The waste is encapsulated in highly durable materials to prevent groundwater from coming into contact with the waste glass while the radioactivity of the waste is still high. 5) The canisters are emplaced in a final repository which is built at a depth of 500 m in rock of low permeability. 6) All tunnels and shafts are filled with a mixture of clay and sand of low permeability. A detailed analysis of possible harmful effects resulting from normal acitivties and from conceivable accidents is presented in a special section. (author)

  2. Nuclear safety philosophy and its general application to fuel management and handling - a regulator's viewpoint

    International Nuclear Information System (INIS)

    Petty, I.C.

    1995-01-01

    The Nuclear Safety Division (NSD) of the Health and Safety Executive (HSE) informs the UK Nuclear Industry of the principles that it applies in assessing whether licensees have demonstrated that their nuclear plants are as safe as is reasonably practicable. The paper commences with a discussion of the non-prescriptive approach to health and safety regulation which is the basis of the regulatory activities of NSD's operating arm -the Nuclear Installations Inspectorate (NII). It then describes in broad terms the overall approach used by NII for analysing the safety of nuclear plant, including fuel, which will cover both deterministic and probabilistic methodologies. The paper then introduces the sections of the Safety Assessment Principles which apply to nuclear fuel safety (both fuel handling and management). Most of these principles are of a general nature and do not just apply to fuel. The paper explains how safety cases might relate to the SAPs and offers some views on how a licensee might interpret them in developing his safety case. Particular emphasis is placed on the importance of submitting a high quality safety case and the type of information that should be in it. The advantages of the approach proposed, to the licensee as well as to the regulator, are identified. (author)

  3. Visual imagery and the user model applied to fuel handling at EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.

    1995-06-01

    The material presented in this paper is based on two studies involving visual display designs and the user`s perspective model of a system. The studies involved a methodology known as Neuro-Linguistic Programming (NLP), and its use in expanding design choices which included the ``comfort parameters`` and ``perspective reality`` of the user`s model of the world. In developing visual displays for the EBR-II fuel handling system, the focus would be to incorporate the comfort parameters that overlap from each of the representation systems: visual, auditory and kinesthetic then incorporate the comfort parameters of the most prominent group of the population, and last, blend in the other two representational system comfort parameters. The focus of this informal study was to use the techniques of meta-modeling and synesthesia to develop a virtual environment that closely resembled the operator`s perspective of the fuel handling system of Argonne`s Experimental Breeder Reactor - II. An informal study was conducted using NLP as the behavioral model in a v reality (VR) setting.

  4. Visual imagery and the user model applied to fuel handling at EBR-II

    International Nuclear Information System (INIS)

    Brown-VanHoozer, S.A.

    1995-01-01

    The material presented in this paper is based on two studies involving visual display designs and the user's perspective model of a system. The studies involved a methodology known as Neuro-Linguistic Programming (NLP), and its use in expanding design choices which included the ''comfort parameters'' and ''perspective reality'' of the user's model of the world. In developing visual displays for the EBR-II fuel handling system, the focus would be to incorporate the comfort parameters that overlap from each of the representation systems: visual, auditory and kinesthetic then incorporate the comfort parameters of the most prominent group of the population, and last, blend in the other two representational system comfort parameters. The focus of this informal study was to use the techniques of meta-modeling and synesthesia to develop a virtual environment that closely resembled the operator's perspective of the fuel handling system of Argonne's Experimental Breeder Reactor - II. An informal study was conducted using NLP as the behavioral model in a v reality (VR) setting

  5. Safety and availability of the fuel handling system at Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Santaliz, Jorge O.; Paredes, Juan A.

    1998-01-01

    The paper attempts the Fuel Handling (F/H) System maintenance and operating methodology at the Embalse Power Station. It doesn't refer to the F/H process, because it's common and well known by all the CANDU Stations. Instead of that, the presentation will be focused on people qualification, training and selection. Also the key subjects for a smooth and successful operation. Additionally will be remarked the human aspect and the role of the person in the organization. The safe and reliable operation of the CNE Fuel Handling System has been always target, supported by the operational experience. The accountability and fitness for the job were the main qualification for the crew members. They have very clear their role and the importance of equipment which they are operating or manipulating. The person who has greater experience and responsibility must struggle continuously to keep the safe and confident operation. Also we have to increase permanently our knowledge with a greater training and experience exchange with another CANDU 6 Station, like this Conference which let us to grow as persons and technicians. It also allows our utility to have access to other realities and work methods. (authors)

  6. Handling of TRIGA spent fuel at the Medical University of Hanover and its return to the United States

    International Nuclear Information System (INIS)

    Hampel, Gabriele; Harke, Heinrich; Kelm, Wieland; Klaus, Uwe

    2008-01-01

    The Medical University of Hannover (MHH) was taking part in the US Department of Energy's (DOE) 'Research Reactor Spent Fuel Acceptance Program' to return its 76 spent TRIGA fuel elements to the United States in the middle of 1999. The fuel elements have been moved to the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. This paper describes the technical facilities for handling the fuel elements at the MHH and the various steps in removing the fuel elements from the reactor, transferring them to the transport cask and shipping them to the INEEL. (authors)

  7. Optimization of fuel exchange machine operation for boiling water reactors using an artificial intelligence technique

    International Nuclear Information System (INIS)

    Sekimizu, K.; Araki, T.; Tatemichi, S.I.

    1987-01-01

    Optimization of fuel assembly exchange machine movements during periodic refueling outage is discussed. The fuel assembly movements during a fuel shuffling were examined, and it was found that the fuel assembly movements consist of two different movement sequences;one is the ''PATH,'' which begins at a discharged fuel assembly and terminates at a fresh fuel assembly, and the other is the ''LOOP,'' where fuel assemblies circulate in the core. It is also shown that fuel-loading patterns during the fuel shuffling can be expressed by the state of each PATH, which is the number of elements already accomplished in the PATH actions. Based on this fact, a scheme to determine a fuel assembly movement sequence within the constraint was formulated using the artificial intelligence language PROLOG. An additional merit to the scheme is that it can simultaneously evaluate fuel assembly movement, due to the control rods and local power range monitor exchange, in addition to normal fuel shuffling. Fuel assembly movements, for fuel shuffling in a 540-MW(electric) boiling water reactor power plant, were calculated by this scheme. It is also shown that the true optimization to minimize the fuel exchange machine movements would be costly to obtain due to the number of alternatives that would need to be evaluated. However, a method to obtain a quasi-optimum solution is suggested

  8. Bacterial foraging-optimized PID control of a two-wheeled machine with a two-directional handling mechanism.

    Science.gov (United States)

    Goher, K M; Fadlallah, S O

    2017-01-01

    This paper presents the performance of utilizing a bacterial foraging optimization algorithm on a PID control scheme for controlling a five DOF two-wheeled robotic machine with two-directional handling mechanism. The system under investigation provides solutions for industrial robotic applications that require a limited-space working environment. The system nonlinear mathematical model, derived using Lagrangian modeling approach, is simulated in MATLAB/Simulink ® environment. Bacterial foraging-optimized PID control with decoupled nature is designed and implemented. Various working scenarios with multiple initial conditions are used to test the robustness and the system performance. Simulation results revealed the effectiveness of the bacterial foraging-optimized PID control method in improving the system performance compared to the PID control scheme.

  9. How problems of storing waste nuclear fuel are handled in some countries

    International Nuclear Information System (INIS)

    Langhe, R.

    1983-01-01

    This report is a survey of the situation in a number of European countries, in the United States and the Soviet Union as well. In all democratic countries, the nuclear power issue is controversial. Everywhere it has met with opposition and criticism, even in countries where nuclear power is officially promoted. Which of the elements comprised in the nuclear power issue is regarded as most controversial varies from country to country. In some countries, final storage and handling of waste nuclear fuel are referred to this category, in others nuclear power plant safety is claimed to be of greater importance. In the last few months, some public opinion has been coupling the peaceful use of nuclear power with nuclear weapons, thereby deeming the greatest danger to be the risk of unwanted distribution of nuclear weapons. Technical difficulties as well as public opinion have indefinitely adjourned the final solution of the disposal of waste nuclear fuel. This problem is of such magnitude that a final solution is urgently needed. Apart from opinions, the existence of waste nuclear power fuel emitting dangerous radiation for over 40 generations to come, makes it a moral obligation to find a way to spare future generations that heritage. (author)

  10. Radiological protection when handling plutonium in a laboratory for experimental fuels

    International Nuclear Information System (INIS)

    Fraser, D.C.

    1978-01-01

    The laboratory for experimental fuels at AEE Winfrith is a small but adaptable workshop capable of fabricating uranium and plutonium as metal or oxide into a variety of fuel elements including pins, plates and coated particle compacts for reactor physics experiments. Experience gained over fifteen years operation has shown that the external radiation dose received by operators, which arises mainly from low energy gamma and X radiation, can be controlled by the widespread use of simple shielding. The radiation from higher energy neutrons cannot be effectively shielded in simple non-automated plant and it becomes more important if large batches of fuel are handled. Inhalation of plutonium oxide is the potentially most important radiological problem. Normally airborne levels of PuO 2 are insignificant; occasionally very high but localised concentrations of airborne material have arisen in working areas, chiefly from minor damage to the flexible part of the containment system, i.e. the gloves and posting bags in glove boxes. Methods employed to measure radiation and inhalation exposure are described and the implications discussed. A fully integrated biological monitoring, in vivo counting and record system is used to ensure that the best estimate of intake is computed for each individual who may be exposed. (author)

  11. Stress analysis of secondary ramp and secondary tilting mechanism of inclined fuel transfer machine for 500 MWe PFBR

    International Nuclear Information System (INIS)

    Prabhakaran, K.M.; Vaze, K.K.; Ghosh, A.K.; Rai, Somesh; Sundarani, A.R.; Patel, R.J.; Agrawal, R.G.

    2004-10-01

    Inclined Fuel Transfer Machine (IFTM) is one of the important machine of the fuel handling system of 500 MWe Prototype Fast Breeder Reactor (PFBR). It is used to transfer core sub-assemblies (CSA) from reactor vessel to fuel building and vice-versa. Secondary ramp and Secondary tilting mechanism (SR/STM) is a part of IFTM which acts as a passage to transfer CSA. This mechanism and components were designed by the Refuelling Technology Division of BARC as per the ASME design code as class 2 component. Being critical in nature and complicated in geometry it was required to check the design of these components by detailed finite element analysis. The loading considered in the present study was static, thermal and seismic conditions. This was done using FEM software COSMOS/M. The Stresses were categorised as per the requirement of the ASME code for various levels of loading (Level A, B and C). Based on the analysis performed, it was concluded that the SR/STM qualifies the requirement of ASME code Section-III NC (Class-2 components). This report gives the details of the studies performed. (author)

  12. Process Knowledge Summary Report for Materials and Fuels Complex Contact-Handled Transuranic Debris Waste

    Energy Technology Data Exchange (ETDEWEB)

    R. P. Grant; P. J. Crane; S. Butler; M. A. Henry

    2010-02-01

    This Process Knowledge Summary Report summarizes the information collected to satisfy the transportation and waste acceptance requirements for the transfer of transuranic (TRU) waste between the Materials and Fuels Complex (MFC) and the Advanced Mixed Waste Treatment Project (AMWTP). The information collected includes documentation that addresses the requirements for AMWTP and the applicable portion of their Resource Conservation and Recovery Act permits for receipt and treatment of TRU debris waste in AMWTP. This report has been prepared for contact-handled TRU debris waste generated by the Idaho National Laboratory at MFC. The TRU debris waste will be shipped to AMWTP for purposes of supercompaction. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU debris waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for waste originating from MFC.

  13. Combining machine learning and ontological data handling for multi-source classification of nature conservation areas

    Science.gov (United States)

    Moran, Niklas; Nieland, Simon; Tintrup gen. Suntrup, Gregor; Kleinschmit, Birgit

    2017-02-01

    Manual field surveys for nature conservation management are expensive and time-consuming and could be supplemented and streamlined by using Remote Sensing (RS). RS is critical to meet requirements of existing laws such as the EU Habitats Directive (HabDir) and more importantly to meet future challenges. The full potential of RS has yet to be harnessed as different nomenclatures and procedures hinder interoperability, comparison and provenance. Therefore, automated tools are needed to use RS data to produce comparable, empirical data outputs that lend themselves to data discovery and provenance. These issues are addressed by a novel, semi-automatic ontology-based classification method that uses machine learning algorithms and Web Ontology Language (OWL) ontologies that yields traceable, interoperable and observation-based classification outputs. The method was tested on European Union Nature Information System (EUNIS) grasslands in Rheinland-Palatinate, Germany. The developed methodology is a first step in developing observation-based ontologies in the field of nature conservation. The tests show promising results for the determination of the grassland indicators wetness and alkalinity with an overall accuracy of 85% for alkalinity and 76% for wetness.

  14. Human machine interface to manually drive rhombic like vehicles in remote handling operations

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Pedro; Vale, Alberto [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ventura, Rodrigo [Institute for Systems and Robotics, Instituto Superior Tecnico, Universidade de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2015-07-01

    In the thermonuclear experimental reactor ITER, a vehicle named CTS is designed to transport a container with activated components inside the buildings. In nominal operations, the CTS is autonomously guided under supervision. However, in some unexpected situations, such as in rescue and recovery operations, the autonomous mode must be overridden and the CTS must be remotely guided by an operator. The CTS is a rhombic-like vehicle, with two drivable and steerable wheels along its longitudinal axis, providing omni-directional capabilities. The rhombic kinematics correspond to four control variables, which are difficult to manage in manual mode operation. This paper proposes a Human Machine Interface (HMI) to remotely guide the vehicle in manual mode. The proposed solution is implemented using a HMI with an encoder connected to a micro-controller and an analog 2-axis joystick. Experimental results were obtained comparing the proposed solution with other controller devices in different scenarios and using a software platform that simulates the kinematics and dynamics of the vehicle. (authors)

  15. Human machine interface to manually drive rhombic like vehicles in remote handling operations

    International Nuclear Information System (INIS)

    Lopes, Pedro; Vale, Alberto; Ventura, Rodrigo

    2015-01-01

    In the thermonuclear experimental reactor ITER, a vehicle named CTS is designed to transport a container with activated components inside the buildings. In nominal operations, the CTS is autonomously guided under supervision. However, in some unexpected situations, such as in rescue and recovery operations, the autonomous mode must be overridden and the CTS must be remotely guided by an operator. The CTS is a rhombic-like vehicle, with two drivable and steerable wheels along its longitudinal axis, providing omni-directional capabilities. The rhombic kinematics correspond to four control variables, which are difficult to manage in manual mode operation. This paper proposes a Human Machine Interface (HMI) to remotely guide the vehicle in manual mode. The proposed solution is implemented using a HMI with an encoder connected to a micro-controller and an analog 2-axis joystick. Experimental results were obtained comparing the proposed solution with other controller devices in different scenarios and using a software platform that simulates the kinematics and dynamics of the vehicle. (authors)

  16. Some information about the radiological protection concerning the TRIGA spent fuel handling at the Medical University of Hanover

    International Nuclear Information System (INIS)

    Hampel, Gabriele; Harke, Heinrich; Klaus, Uwe; Loercher, Gunther

    2008-01-01

    The Medical University of Hanover (MHH) returned its 76 spent TRIGA fuel elements to the United States in summer of 1999. For the transportation inside the MHH control areas were installed outside the reactor area, along the transfer route in the department of nuclear medicine and in the temporary building. During fuel handling at MHH a lot of radiation protection measures were necessary. This paper presents methods and results of the radiological protection measurements. (authors)

  17. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  18. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    International Nuclear Information System (INIS)

    Peterson, Per; Greenspan, Ehud

    2015-01-01

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3 . This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel

  19. Proposed master-slave and automated remote handling system for high-temperature gas-cooled reactor fuel refabrication

    International Nuclear Information System (INIS)

    Grundmann, J.G.

    1974-01-01

    The Oak Ridge National Laboratory's Thorium-Uranium Recycle Facility (TURF) will be used to develop High-Temperature Gas-Cooled Reactor (HTGR) fuel recycle technology which can be applied to future HTGR commercial fuel recycling plants. To achieve recycle capabilities it is necessary to develop an effective material handling system to remotely transport equipment and materials and to perform maintenance tasks within a hot cell facility. The TURF facility includes hot cells which contain remote material handling equipment. To extend the capabilities of this equipment, the development of a master-slave manipulator and a 3D-TV system is necessary. Additional work entails the development of computer controls to provide: automatic execution of tasks, automatic traverse of material handling equipment, automatic 3D-TV camera sighting, and computer monitoring of in-cell equipment positions to prevent accidental collisions. A prototype system which will be used in the development of the above capabilities is presented. (U.S.)

  20. Preliminary design for spent fuel canister handling systems in a canister transfer and installation vehicle

    International Nuclear Information System (INIS)

    Wendelin, T.; Suikki, M.

    2008-12-01

    The report presents a spent fuel canister transfer and installation vehicle. The vehicle is used for carrying the fuel canister into a disposal tunnel and installing it into a deposition hole. The report outlines basic requirements and a design for canister handling equipment used in a canister transfer and installation vehicle, a description regarding the operation and maintenance of the equipment, as well as a cost estimate. Specific vehicles will be manufactured for all canister types in order to minimize the height of the disposal tunnels. This report is only focused on a transfer and installation vehicle for OL1-2 fuel canisters. Detailed designing and selection of final components have not yet been carried out. The report also describes the vehicle's requirements for the structures of a repository system, as well as actions in possible malfunction or fault situations. The spent fuel canister is brought from an encapsulation plant by a canister lift down to the repository level. The fuel canister is driven from the canister lift by an automated guided vehicle onto a canister hoist at a canister loading station. The canister transfer and installation vehicle is waiting for the canister with its radiation shield in an upright position above the canister hoist. The hoist carries the canister upward until the vehicle's own lifting means grab hold of the canister and raise it up into the vehicle's radiation shield. This is followed by turning the radiation shield to a transport position and by closing it in a radiation-proof manner against a rear radiation shield. The vehicle is driven along the central tunnel into the disposal tunnel and parked on top of the deposition hole. The vehicle's radiation shield is turned to the upright position and the canister is lowered with the vehicle's hydraulic winches into a bentonite-lined deposition hole. The radiation shield is turned back to the transport position and the vehicle can be driven out of the disposal tunnel

  1. Design Evaluation of UIS and In-vessel Fuel Transfer Machine for a 1200MWe SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Kim, Seok Hoon; Park, Chang Gyu; Lee, Su Yeon

    2008-11-15

    The report describes the structural applicability of the upper internal structure (UIS) and the in-vessel fuel transfer machine for a 1200MWe sodium cooled fast reactor (SFR) of a pool type. In the conceptual design, a two rotating plug type as a refueling system is considered. For the two rotating plug type, the diameters of large and small rotating plugs are determined by 7.2m and 5.6m, respectively. Through the use of an inner fuel transfer machine and the lift change machine with a fixed arm length of 1.10m installed on a small rotating plug, all the core assemblies are accessed within 7mm accuracy. The UIS diameter is determined by 4.7m, which includes the all control drive lines in upper part, the diameter of UIS lower part is restricted by 4.4 m to secure the rotation angle of a refueling machine.

  2. Automatic fuel charging machine for nuclear power plants

    International Nuclear Information System (INIS)

    Neda, Toshikazu; Aoki, Shigeo.

    1978-01-01

    Purpose: To enable smooth, safety and rapid refueling by automatically conducting a series of fuel exchange steps through the combination of calculations for fuel control and refueling works. Constitution: Processing operations are conducted based on various data from the memory unit of a computer that processes and stores various data inputted from a nuclear power plant, calculation systems stored in the memory unit of another computer, peripheral units such as typewriters and process input units. A refueling platform is operated by way of a platform control device and a platform driving device, and fuel exchange is conducted by the operation of a channel mounting and demounting device. (Yoshino, Y.)

  3. Support vector machine to predict diesel engine performance and emission parameters fueled with nano-particles additive to diesel fuel

    Science.gov (United States)

    Ghanbari, M.; Najafi, G.; Ghobadian, B.; Mamat, R.; Noor, M. M.; Moosavian, A.

    2015-12-01

    This paper studies the use of adaptive Support Vector Machine (SVM) to predict the performance parameters and exhaust emissions of a diesel engine operating on nanodiesel blended fuels. In order to predict the engine parameters, the whole experimental data were randomly divided into training and testing data. For SVM modelling, different values for radial basis function (RBF) kernel width and penalty parameters (C) were considered and the optimum values were then found. The results demonstrate that SVM is capable of predicting the diesel engine performance and emissions. In the experimental step, Carbon nano tubes (CNT) (40, 80 and 120 ppm) and nano silver particles (40, 80 and 120 ppm) with nanostructure were prepared and added as additive to the diesel fuel. Six cylinders, four-stroke diesel engine was fuelled with these new blended fuels and operated at different engine speeds. Experimental test results indicated the fact that adding nano particles to diesel fuel, increased diesel engine power and torque output. For nano-diesel it was found that the brake specific fuel consumption (bsfc) was decreased compared to the net diesel fuel. The results proved that with increase of nano particles concentrations (from 40 ppm to 120 ppm) in diesel fuel, CO2 emission increased. CO emission in diesel fuel with nano-particles was lower significantly compared to pure diesel fuel. UHC emission with silver nano-diesel blended fuel decreased while with fuels that contains CNT nano particles increased. The trend of NOx emission was inverse compared to the UHC emission. With adding nano particles to the blended fuels, NOx increased compared to the net diesel fuel. The tests revealed that silver & CNT nano particles can be used as additive in diesel fuel to improve complete combustion of the fuel and reduce the exhaust emissions significantly.

  4. Testing a CANDU-fueling machine at the Institute for Nuclear Research Pitesti

    International Nuclear Information System (INIS)

    Cojocaru, Virgil

    2006-01-01

    In 2003, as a national and European premiere, the Fueling Machine Head no. 4 (F/M) for the Nuclear Power Plant Cernavoda Unit 2 (NPP) was successfully tested at the Institute for Nuclear Research Pitesti (INR). In 2005, the second Fueling Machine (no. 5) has tested for the Nuclear Power Plant Cernavoda Unit 2. The Institute's main objective is to develop scientific and technological support for the Romanian Nuclear Power Program. Testing the Fueling Machines at INR Pitesti is part of the overall program to assimilate the CANDU technology in Romania. To perform the tests of these machines at INR Pitesti, a special testing rig has built being available for this goal. Both the testing rig and staff had successfully assessed by the AECL representatives during two missions. There was a delivery contract between GEC Canada and Nuclear Power Plant Cernavoda - Unit 2 to provide the Fueling Machines no. 4 and no. 5 in Romania before testing activity. As a first conclusion, the Institute for Nuclear Research Pitesti has the facilities, the staff and the experience to perform possible co-operations with any CANDU Reactor owner

  5. Design method of control system for HTGR fuel handling process with control Petri net

    International Nuclear Information System (INIS)

    Han Zandong; Luo Sheng; Liu Jiguo

    2008-01-01

    As a complex mechanical system,the fuel handling system (FHS) of pebble-bed high temperature gas cooled reactor (HTGR) is with the features of complicated structure, numerous control devices and strict working scheduling. It is very important to precisely describe the function of FHS and effectively design its control system. A design method of control system based on control Petri net (CPN) is introduced in this paper. By associating outputs and operations with places, associating inputs and conditions with transitions, and introducing macro-places and macro-actions, the CPN realizes hierarchy design of complex control system. Based on the analysis of basic functions and working flow of FHS, its control system is described and designed by CPN. According to the firing regulation of transition,the designed CPN can be easily converted into LAD program of PLC, which can be implemented on the FHS simulating control test-bed. Application illuminates that proposed method has the advantages of clear design structure, exact description power and effective design ability of control program, which can meet the requirements of FHS control sys-tem design. (authors)

  6. Generic environmental impact statement on handling and storage of spent light water power reactor fuel. Appendices

    International Nuclear Information System (INIS)

    1978-03-01

    Detailed appendices are included with the following titles: light water reactor fuel cycle, present practice, model 1000MW(e) coal-fired power plant, increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data (1976-2000), characteristics of nuclear fuel, and ''away-from-reactor'' storage concept

  7. Design Improvements of the Capsule Components and the Handling Tools for an Effective Utilization of the Capsule Assembly Machine

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Kim, Jong Kium; Youm, Ki Un; Yoon, Ki Byung; Choi, Myung Hwan; Kim, Bong Gu

    2006-01-01

    Various in-pile test programs for the development of new fuels and materials are planned for the HANARO reactor. To meet the demands for the required tests in the HANARO reactor, new capsule assembling technology is required in the HANARO reactor. For this reason, a series of analyses and experiments was performed in 2005. For the assembly workability tests of the capsule components, three different kinds of protection tubes and two different shapes of the locking bolt heads were proposed and tested. It was confirmed that the newly designed protection tube and bolts worked quite well without any problems. Since the new structure is quite similar to that of the currently used capsule, it was assumed that an additional vibration tests and seismic analysis would not be needed. Through the stress analysis of the three proposed structures by using ANSYS code, it showed that the maximum displacement and stress intensity for the tube reducer were 1.57mm and 21MPa, respectively. To improve the workability and handling capability of the bolting and clamping tools of stainless steel 304, Al6061/T6 was selected as one of the candidates and thus new tools were manufactured and tested. The assembly test results showed that the new tools were found to be useful for executing key tasks such as a bolting and a clamping and they were much faster than the old tools made of stainless steel, thereby increasing the workability rate and lowering the manufacturing costs

  8. Proceedings of the 6. international conference on stability and handling of liquid fuels. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Giles, H.N. [ed.] [Deputy Assistant Secretary for Strategic Petroleum Reserve, Washington, DC (United States). Operations and Readiness Office

    1998-12-01

    Volume 1 of these proceedings contain 29 papers related to aviation fuels and long term and strategic storage. Studies investigated fuel contamination, separation processes, measurement techniques, thermal stability, compatibility with fuel system materials, oxidation reactions, and degradation during storage.

  9. Evaluation of Multi Canister Overpack (MCO) Handling Machine Uplift Restraint for a Seismic Event During Repositioning Operations

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    Insertion of the Multi-Canister Overpack (MCO) assemblies into the Canister Storage Building (CSB) storage tubes involves the use of the MCO Handling Machine (MHM). During MCO storage tube insertion operations, inadvertent movement of the MHM is prevented by engaging seismic restraints (''active restraints'') located adjacent to both the bridge and trolley wheels. During MHM repositioning operations, the active restraints are not engaged. When the active seismic restraints are not engaged, the only functioning seismic restraints are non-engageable (''passive'') wheel uplift restraints which function only if the wheel uplift is sufficient to close the nominal 0.5-inch gap at the uplift restraint interface. The MHM was designed and analyzed in accordance with ASME NOG-1-1995. The ALSTHOM seismic analysis reported seismic loads on the MHM uplift restraints and EDERER performed corresponding structural calculations to demonstrate structural adequacy of the seismic uplift restraint hardware. The ALSTHOM and EDERER calculations were performed for a parked MHM with the active seismic restraints engaged, resulting in uplift restraint loading only in the vertical direction. In support of development of the CSB Safety Analysis Report (SAR), an evaluation of the MHM seismic response was requested for the case where the active seismic restraints are not engaged. If a seismic event occurs during MHM repositioning operations, a moving contact at a seismic uplift restraint would introduce a friction load on the restraint in the direction of the movement. These potential horizontal friction loads on the uplift restraints were not included in the existing restraint hardware design calculations. One of the purposes of the current evaluation is to address the structural adequacy of the MHM seismic uplift restraints with the addition of the horizontal friction associated with MHM repositioning movements

  10. Expert system for assisting the diagnostic and localisation of breakdowns on the fuel elements loading machine

    International Nuclear Information System (INIS)

    Merlin, J.; Pradal, B.

    1990-01-01

    An expert system is developed in order to minimize the time lost through breakdowns of the fuel loading device. The expert system developed by FRAMATOME uses MAINTEX software. The expert systems MACHA and SEDMAC were designed respectively for use on 1300 MWe and 900 MWe loading machines [fr

  11. Evolution of the design of fuel handling control system in 220 MWe Indian PHWRs

    International Nuclear Information System (INIS)

    Dhruvanarayana, L.; Gupta, H.; Bharathkumar, M.

    1996-01-01

    Following two CANDU type reactors at Rajasthan (RAPS-1 and 2), three nuclear power stations, each of two units of 220 MWe has been in operation at Rajasthan (RAPS-1 and 2). Madras (MAPS-1 and 2). Narora (NAPS-1 and 2) and Kakrapar (KAPS-1 and 2). Two more stations, also of 220 MWe capacity, are under construction at Rajasthan (RAPP-3 and 4) and Kaiga (Kaiga-1 and 2). These are natural uranium fuelled pressurized heavy water cooled and heavy water moderated reactors (PHWRs). The two units at Rajasthan viz RAPS-1 and 2, were built with the technical collaboration with Canada, and the rest of the units have been designed and built indigenously, incorporating a number of modifications, particularly in the on-power refuelling system. The evolution of the design of the Fuel Handling Control systems of these reactors, taking into consideration operational needs, safety aspects and maintainability are highlighted in this paper. A combination of hydraulic and electronic control has been provided to enable the operations. In RAPS-1 and 2, hardwired electronic controls were provided, while in MAPS-1 and 2, the hardwired system was improved. From NAPS onwards, a computerized control system with hardwired interlock logic has been provided. New devices like coarse-fine potentiometers, special oil filled potentiometer assembly, rectilinear potentiometers etc., were specified from NAPS onwards. Positioning logic is computerized providing flexibility and expendability. Digital panel meters and indicating lamps have been provided for manual mode operations, while CRT (cathode-ray tube) monitors help in computer mode operations. Hydraulic controls which comprise D 2 0 hydraulics, H 2 0 hydraulics and oil hydraulics have been improved from NAPS onwards. Hydraulic panels have been relocated in accessible areas to reduce radiation doses and for better maintainability. All electric drives including X and Y drives were modified as hydraulic drives for better control. New types of valves

  12. Analysis of fuel-handling incidents (safety analysis detailed report no. 5). PEC Brasimone reactor design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-15

    The features covered by this report deal with the equipment and cells in which the handling, examination, measurement, conditioning and storage of core elements are carried out. The operations covered range from the receiving of new element shipments to their insertion in the vessel (excluding handling inside the vessel itself, which is covered in report no. 2) and removal of the spent-elements from the vessel, transfer to their final storage and their ultimate loading into containers for transport outside the plant. The incident analysis along the path of the spent fuel was conducted with the same method adopted for other plant systems. It is treated separately here because the operation of the handling system is practically autonomous from reactor operation.

  13. TMI-2 [Three Mile Island Nuclear Power Station] fuel canister and core sample handling equipment used in INEL hot cells

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Shurtliff, W.T.; Lynch, R.J.; Croft, K.M.; Whitmill, L.J.; Allen, S.M.

    1987-01-01

    This paper describes the specialized remote handling equipment developed and used at the Idaho National Engineering Laboratory (INEL) to handle samples obtained from the core of the damaged Unit 2 reactor at Three Mile Island Nuclear Power Station (TM-2). Samples of the core were removed, placed in TMI-2 fuel canisters, and transported to the INEL. Those samples will be examined as part of the analysis of the TMI-2 accident. The equipment described herein was designed for removing sample materials from the fuel canisters, assisting with initial examination, and processing samples in preparation for detailed examinations. The more complex equipment used microprocessor remote controls with electric motor drives providing the required force and motion capabilities. The remaining components were unpowered and manipulator assisted

  14. Operational analysis and improvement of a spent nuclear fuel handling and treatment facility using discrete event simulation

    International Nuclear Information System (INIS)

    Garcia, H.E.

    2000-01-01

    Spent nuclear fuel handling and treatment often require facilities with a high level of operational complexity. Simulation models can reveal undesirable characteristics and production problems before they become readily apparent during system operations. The value of this approach is illustrated here through an operational study, using discrete event modeling techniques, to analyze the Fuel Conditioning Facility at Argonne National Laboratory and to identify enhanced nuclear waste treatment configurations. The modeling approach and results of what-if studies are discussed. An example on how to improve productivity is presented.

  15. Computerised weld strength testing machine for PHWR fuel elements with a versatile control system

    International Nuclear Information System (INIS)

    Gupta, U.C.; Sastry, V.S.; Rasheed, Jawad; Bibawe, S.R.

    1994-01-01

    Spacer pads and bearing pads are resistance spot welded on PHWR fuel elements to ensure inter-element gap for coolant flow. These welds are subjected to destructive tests as per SQC specifications while qualifying a machine and during production. The testing machine used earlier over the years was tedious involving manual operations of clamping, tool actuation, increasing and decreasing the pressure, referring to charts and statistical calculations. To carry out the destructive testing of the welds conveniently and reliably, an automatic machine is developed in-house in which are incorporated a quartz force transducer and a computerized data-acquisition and processing system together with a very versatile electronic control system based on a single-chip microcomputer. This paper describes the salient features of the machine and the control system. (author). 4 figs

  16. Evolution of the design of fuel handling control system in 220 MWe Indian PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Dhruvanarayana, L; Gupta, H; Bharathkumar, M [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    Following two CANDU type reactors at Rajasthan (RAPS-1 and 2), three nuclear power stations, each of two units of 220 MWe has been in operation at Rajasthan (RAPS-1 and 2). Madras (MAPS-1 and 2). Narora (NAPS-1 and 2) and Kakrapar (KAPS-1 and 2). Two more stations, also of 220 MWe capacity, are under construction at Rajasthan (RAPP-3 and 4) and Kaiga (Kaiga-1 and 2). These are natural uranium fuelled pressurized heavy water cooled and heavy water moderated reactors (PHWRs). The two units at Rajasthan viz RAPS-1 and 2, were built with the technical collaboration with Canada, and the rest of the units have been designed and built indigenously, incorporating a number of modifications, particularly in the on-power refuelling system. The evolution of the design of the Fuel Handling Control systems of these reactors, taking into consideration operational needs, safety aspects and maintainability are highlighted in this paper. A combination of hydraulic and electronic control has been provided to enable the operations. In RAPS-1 and 2, hardwired electronic controls were provided, while in MAPS-1 and 2, the hardwired system was improved. From NAPS onwards, a computerized control system with hardwired interlock logic has been provided. New devices like coarse-fine potentiometers, special oil filled potentiometer assembly, rectilinear potentiometers etc., were specified from NAPS onwards. Positioning logic is computerized providing flexibility and expendability. Digital panel meters and indicating lamps have been provided for manual mode operations, while CRT (cathode-ray tube) monitors help in computer mode operations. Hydraulic controls which comprise D{sub 2}0 hydraulics, H{sub 2}0 hydraulics and oil hydraulics have been improved from NAPS onwards. Hydraulic panels have been relocated in accessible areas to reduce radiation doses and for better maintainability. All electric drives including X and Y drives were modified as hydraulic drives for better control. New types of

  17. Two CANDU fueling machines tested at the Institute For Nuclear Research - Pitesti

    International Nuclear Information System (INIS)

    Doca, Cezar; Cojocaru, Virgil

    2005-01-01

    In 2003, as a national and European premiere, at the Institute for Nuclear Research Pitesti (INR), the Fueling Machine Head no.4 (F/M) for the Nuclear Power Plant Cernavoda - Unit 2 was successfully tested. In 2005, a second Fueling Machine (no.5) was tested for the Nuclear Power Plant Cernavoda - Unit 2. The Institute's main objective is to develop scientific and technological support for the Romanian Nuclear Power Program. Testing the Fueling Machines at INR Pitesti is part of the overall program to assimilate in Romania the CANDU technology. To perform the tests of these machines at INR Pitesti, a special testing rig was built and is available for this goal. Both the testing rig and staff had successfully assessed by the AECL representatives during two missions. There was a delivery contract between GEC Canada and Nuclear Power Plant Cernavoda - Unit 2 to provide the Fueling Machines no. 4 and no. 5 in Romania before testing operation. As a first conclusion, the Institute for Nuclear Research Pitesti has the facilities, the staff and the experience to perform possible co-operations with any other CANDU Reactor owner. This experience will support the next steps concerning F/M commissioning in the NPP Cernavoda - Unit 2 and also give the confidence to the end-users that the Institute's team can provide technical assistance during the operation. Also, the obtained results demonstrate that the overall refurbishment of the F/M control system in Unit 1 and Unit 2 will be possible. The paper presents: - a short description of the F/M head;- a short description of the F/M test rig; - the computer control system; - the F/M testing activities; -results and expectations. (authors)

  18. Two CANDU fueling machines tested at the Institute For Nuclear Research - Pitesti

    International Nuclear Information System (INIS)

    Doca, C.; Cojocaru, V.

    2005-01-01

    Full text: In 2003, as a national and European premiere, at the Institute for Nuclear Research Pitesti (INR), the Fueling Machine Head no.4 (F/M) for the Nuclear Power Plant Cernavoda - Unit 2 was successfully tested. In 2005, a second Fueling Machine (no.5) was tested for the Nuclear Power Plant Cernavoda - Unit 2. The Institute's main objective is to develop scientific and technological support for the Romanian Nuclear Power Program. Testing the Fueling Machines at INR Pitesti is part of the overall program to assimilate in Romania the CANDU technology. To perform the tests of these machines at INR Pitesti, a special testing rig was built and is available for this goal. Both the testing rig and staff had successfully assessed by the AECL representatives during two missions. There was a delivery contract between GEC Canada and Nuclear Power Plant Cernavoda - Unit 2 to provide the Fueling Machines no. 4 and no. 5 in Romania before testing operation. As a first conclusion, the Institute for Nuclear Research Pitesti has the facilities, the staff and the experience to perform possible co-operations with any other CANDU Reactor owner. This experience will support the next steps concerning F/M commissioning in the NPP Cernavoda - Unit 2 and also give the confidence to the end-users that the Institute's team can provide technical assistance during the operation. Also, the obtained results demonstrate that the overall refurbishment of the F/M control system in Unit 1 and Unit 2 will be possible. The paper presents: - a short description of the F/M head;- a short description of the F/M test rig; - the computer control system; - the F/M testing activities; -results and expectations. (authors)

  19. Installation and method for handling fuel assemblies of fast nuclear reactors

    International Nuclear Information System (INIS)

    Aubert, Michel; Renaux, Charley.

    1982-01-01

    This invention concerns an installation and a method for handling the assemblies which makes it possible to have a large revolving plug smaller in diameter than that of the presently known solutions. This large, coaxial to the core, revolving plug has a handling arm enabling a fraction of the assemblies to be reached and deposited in a handling well. Through a small offset revolving plug the remainder of the assemblies can be reached and deposited in a pick-up well accessible to the arm of the large revolving plug [fr

  20. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

  1. Proceedings of the 6. international conference on stability and handling of liquid fuels. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Giles, H.N. [ed.] [Deputy Assistant Secretary for Strategic Petroleum Reserve, Washington, DC (United States). Operations and Readiness Office

    1998-12-01

    Volume 2 of these proceedings contain 42 papers arranged under the following topical sections: Fuel blending and compatibility; Middle distillates; Microbiology; Alternative fuels; General topics (analytical methods, tank remediation, fuel additives, storage stability); and Poster presentations (analysis methods, oxidation kinetics, health problems).

  2. Investigation regarding the safety of handling the fuel assemblies for the nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    It was concluded previously that the general inspection of safety and the repair of shielding can be carried out as the fuel assemblies are charged, and the safety can be secured sufficiently. According to the decision by the meeting of cabinet ministers concerned with the nuclear ship ''Mutsu'', the Mutsu General Inspection and Repair Technology Investigation Committee investigated on the basic concept regarding the method and the safety of taking out, transporting and preserving the fuel assemblies. 112 fuel rods and 9 burnable poison rods are arranged into the square grid of 11 x 11 in a fuel assembly, and 32 fuel assemblies are employed. The contents of the investigation are the outline of the fuel assemblies, the present states of nuclear fission products, surface dose rate and soundness of the fuel assemblies, the safety of taking out, transporting and preserving the fuel assemblies, the measures required for securing the safety, and the place for taking out the fuel assemblies. In case of taking out, transporting and preserving the fuel assemblies, it is considered in view of the present state of the fuel assemblies that the safety can be secured sufficiently if the works are carried out carefully by taking the methods and conditions investigated into consideration. Also the committee reached already the conclusion described at the outset. (Kako, I.)

  3. Automated fuel fabrication- a vision comes true

    International Nuclear Information System (INIS)

    Hemantha Rao, G.V.S.; Prakash, M.S.; Setty, C.R.P.; Gupta, U.C.

    1997-01-01

    When New Uranium Fuel Assembly Project at Nuclear Fuel Complex (NFC) begins production, its operator will have equipment provided with intramachine handling systems working automatically by pressing a single button. Additionally simple low cost inter machine handling systems will further help in critical areas. All these inter and intra machine handling systems will result in improved reliability, productivity and quality. The fault diagnostics, mimics and real time data acquisition systems make the plant more operator friendly. The paper deals with the experience starting from layout, selection of product carriers, different handling systems, the latest technology and the integration of which made the vision on automation in fuel fabrication come true. (author)

  4. Spent fuel handling system for a geologic storage test at the Nevada Test Site

    International Nuclear Information System (INIS)

    Duncan, J.E.; House, P.A.; Wright, G.W.

    1980-01-01

    The Lawrence Livermore Laboratory is conducting a test of the geologic storage of encapsulated spent commercial reactor fuel assemblies in a granitic rock at the Nevada Test Site. The test, known as the Spent Fuel Test-Climax (SFT-C), is sponsored by the US Department of Energy, Nevada Operations Office. Eleven pressurized-water-reactor spent fuel assemblies are stored retrievably for three to five years in a linear array in the Climax stock at a depth of 420 m

  5. Removal of spent fuel from the TVR reactor for reprocessing and proposals for the RA reactor spent fuel handling

    International Nuclear Information System (INIS)

    Volkov, E.B.; Konev, V.N.; Shvedov, O.V.; Bulkin, S.Yu; Sokolov, A.V.

    2002-01-01

    The 2,5 MW heavy-water moderated and cooled research reactor TVR was located at the Moscow Institute for Theoretical and Experimental Physics site. In 1990 the final batch of spent nuclear fuel (SNF) from the TVR reactor was transported for reprocessing to Production Association (PA) 'Mayak'. This transportation of the SNF was a part of TVR reactor decommissioning. The special technology and equipment was developed in order to fulfill the preparation of TVR SNF for transportation. The design of the TVR reactor and the fuel elements used are similar to the design and fuel elements of the RA reactor. Two different ways of RA spent fuel elements for transportation to reprocessing plant are considered: in aluminum barrels, and in additional cans. The experience and equipment used for the preparing TVR fuel elements for transportation can help the staff of RA reactor to find the optimal way for these technical operations. (author)

  6. SAF: the next generation process for radiotoxic material handling in the nuclear fuel industry

    International Nuclear Information System (INIS)

    Nyman, D.H.; Graham, R.A.

    1984-01-01

    In 1980 the Secure Automated Fabrication (SAF) Project was established with the goal to design, build, and operate a remote process for manufacturing breeder reactor fuel pins. The SAF line will be housed in the Fuels and Materials Examination Facility (FMEF) at the Hanford site. The fabrication system and supporting operations are designed for computer-controlled operation from a centralized control room. In addition to improved worker protection, remote and automated fuel fabrication operations will result in enhanced safeguards and accountability of fuel material, improved product quality, and increased productivity. Installation of the SAF line equipment has started. Qualification runs are scheduled to begin in 1986 with production commencing in 1987

  7. Dissecting the fuel element replacement machine in the NPP Muehlheim-Kaerlich using a wire saw

    International Nuclear Information System (INIS)

    Keller, Melanie; Ambos, Frank; Platt, Klaus

    2008-01-01

    By the end of 2000 the NPP Muehlheim-Kaerlich is decommissioned and has to be dismantled. The authors describe the demolition of the fuel element replacement machine. The work is aimed to declassify the waste without restraints for conventional disposal. Within 7 weeks the component has to be dissected using a wire saw. The authors describe the concept and its adaptation to the boundary conditions and experiences during the realization of the demolition concept.

  8. Machine-Building for Fuel and Energy Complex: Perspective Forms of Interaction

    Science.gov (United States)

    Nikitenko, S. M.; Goosen, E. V.; Pakhomova, E. A.; Rozhkova, O. V.; Mesyats, M. A.

    2017-10-01

    The article is devoted to the study of the existing forms of cooperation between the authorities, business and science in the fuel and energy complex and the machine-building industry at the regional level. The possibilities of applying the concept of the “triple helix” and its multi-helix modifications for the implementation of the import substitution program for high- tech products have been considered.

  9. EvoBot: An Open-Source, Modular Liquid Handling Robot for Nurturing Microbial Fuel Cells

    DEFF Research Database (Denmark)

    Faina, Andres; Nejatimoharrami, Farzad; Støy, Kasper

    2016-01-01

    makes it difficult to apply conventional liquid handling robots as they are designed to automate a predefined task. In order to address these issues, we have developed an open source liquid handling robot, EvoBot. It uses a modular approach, which gives us the possibility to reconfigure the robot...... for different experiments and make it possible for users to add functionality by just developing a function specific module. In addition, it provides sensors and extra functionality for monitoring an experiment, which allows researchers to perform interactive experiments with the aim of prolonging non...

  10. Handling device for nuclear fuel assemblies and assembly appropriate for such a device

    International Nuclear Information System (INIS)

    Cransac, J.P.; Jaquelin, R.; Renaux, C.

    1985-01-01

    The handling device comprises a guide tube of which axis is vertical, in which a grab moves, hanging from a chain, under the action of a back-geared motor. The grab being stopped in its rotation in the guide tube, an assembly can be gripped with a bayonet system while controlling the rotation of the grab - guide tube system a back-geared motor. The device can be hanged from the small or large rotating plug of a fast neutron reactor. It can be used in a handling flask [fr

  11. State fund of decommissioning of nuclear installations and handling of spent nuclear fuels and nuclear wastes (Slovak Republic)

    International Nuclear Information System (INIS)

    Kozma, Milos

    2006-01-01

    State Fund for Decommissioning of Nuclear Installations and Handling of Spent Nuclear Fuels and Nuclear Wastes was established by the Act 254/1994 of the National Council of the Slovak Republic as a special-purpose fund which concentrates financial resources intended for decommissioning of nuclear installations and for handling of spent nuclear fuels and radioactive wastes. The Act was amended in 2000, 2001 and 2002. The Fund is legal entity and independent from operator of nuclear installations Slovak Power Facilities Inc. The Fund is headed by Director, who is appointed and recalled by Minister of Economy of the Slovak Republic. Sources of the Fund are generated from: a) contributions by nuclear installation operators; b) penalties imposed by Nuclear Regulatory Authority of the Slovak Republic upon natural persons and legal entities pursuant to separate regulation; c) bank credits; d) interest on Fund deposits in banks; e) grants from State Budget; f) other sources as provided by special regulation. Fund resources may be used for the following purposes: a) decommissioning of nuclear installations; b) handling of spent nuclear fuels and radioactive wastes after the termination of nuclear installation operation; c) handling of radioactive wastes whose originator is not known, including occasionally seized radioactive wastes and radioactive materials stemming from criminal activities whose originator is not known, as confirmed by Police Corps investigator or Ministry of Health of the Slovak Republic; d) purchase of land for the establishment of nuclear fuel and nuclear waste repositories; e) research and development in the areas of decommissioning of nuclear installations and handling of nuclear fuels and radioactive wastes after the termination of the operation of nuclear installations; f) selection of localities, geological survey, preparation, design, construction, commissioning, operation and closure of repositories of spent nuclear fuels and radioactive wastes

  12. Proceedings of the 5th international conference on stability and handling of liquid fuels. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Giles, H.N. [ed.

    1995-04-01

    Volume 2 of these proceedings contains 34 papers divided into the following sessions: Deposit and insolubles measurement (5 papers); Gasolines (4 papers); Heavy oils and refinery processing (3 papers); Middle distillate fuels (7 papers); New fuels and environmental mandates (5 papers); and a Poster session (10 papers). Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  13. Method and apparatus for the handling and inspection of a nuclear reactor fuel element

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1975-01-01

    The non-destructive inspection, for instance, of spent fuel elements and their dismantling are carried out under water in a pool. For this purpose, the fuel elements are attached to a bar which can be moved under water from the vertical into horizontal directions by means of a winch. The bar proper is suspended from a bridge spanning the pool. On one side, the bar is pivoted in a pin installed in components suspended from the bridge, whilst the movement of the bar is limited by a horizontal stop. In the vertical position, the fuel elements and components, respectively, such as fuel elements, are taken up and inspected in the horizontal position by means of TV systems or periscopes. The fuel elements are conveyed by a trolley. Dismantling of the fuel elements under water is carried out by special tools, such as cranks and connecting rods which, inter alia, put the individual fuel rods onto grids prior to inspection, disengage the clamps by means of grid disconnecting systems, remove the fuel rods from the grids and put them on the bars. (DG/RF) [de

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

    Energy Technology Data Exchange (ETDEWEB)

    Desbats, Philippe [CEA - Direction de la Recherche Technologique / LIST, BP 6 - 92265, Fontenay-aux-Roses cedex (France); Piolain, Gerard [COGEMA-HAG/DMCO, AREVA NC SA, 2, rue Paul Dautier, BP 4, 78 141 Velizy Cedex (France)

    2006-07-01

    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

  15. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  16. Preparation for commissioning of nuclear plant with reference to British Nuclear Fuels Plc fuel handling plant project

    International Nuclear Information System (INIS)

    Bamber, D.

    1987-01-01

    The new Fuel Handing Plant at British Nuclear Fuels Sellafield is part of a Pound 550M complex which provides facilities for the receipt, storage and mechanical preparation of both magnox and A.G.R. fuel. The plant is very large and complex with considerable use of computer based process control systems, providing for physical and nuclear safety. The preparation of such plant for ''active'' commissioning necessitates a great many physical checks and technical evaluations in support of its safety case. This paper describes arrangements for plant commissioning checks, against the regulatory framework and explains the physical preparations necessary for their timely accomplishment. (author)

  17. Preliminary design and manufacturing feasibility study for a machined Zircaloy triangular pitch fuel rod support system (grids) (AWBA development program)

    International Nuclear Information System (INIS)

    Horwood, W.A.

    1981-07-01

    General design features and manufacturing operations for a high precision machined Zircaloy fuel rod support grid intended for use in advanced light water prebreeder or breeder reactor designs are described. The grid system consists of a Zircaloy main body with fuel rod and guide tube cells machined using wire EDM, a separate AM-350 stainless steel insert spring which fits into a full length T-slot in each fuel rod cell, and a thin (0.025'' or 0.040'' thick) wire EDM machined Zircaloy coverplate laser welded to each side of the grid body to retain the insert springs. The fuel rods are placed in a triangular pitch array with a tight rod-to-rod spacing of 0.063 inch nominal. Two dimples are positioned at the mid-thickness of the grid (single level) with a 90 0 included angle. Data is provided on the effectiveness of the manufacturing operations chosen for grid machining and assembly

  18. An electrical pulse hydride injector (EPHI) for reactor fueling and tritium handling applications

    International Nuclear Information System (INIS)

    Azizov, E.A.; Kareev, Yu.A.; Savotkin, A.N.; Frunze, V.V.; Penzhorn, R.D.; Glugla, M.

    1995-01-01

    An electrical pulse hydride injector (EPHI) has been developed for reactor fuelling as well as for handling of hydrogen isotopes in facilities operating with tritium. Salient features of the EPHI are the accuracy with which the fuelling rate can be controlled and the avoidance of a pressurized ballast. The generator is simple and allows for safe operation with tritium. (orig.)

  19. Impact of thermodynamic properties and heat loss on ignition of transportation fuels in rapid compression machines

    KAUST Repository

    Ahmed, Ahfaz

    2018-01-30

    Rapid compression machines (RCM) are extensively used to study autoignition of a wide variety of fuels at engine relevant conditions. Fuels ranging from pure species to full boiling range gasoline and diesel can be studied in an RCM to develop a better understanding of autoignition kinetics in low to intermediate temperature ranges. In an RCM, autoignition is achieved by compressing a fuel/oxidizer mixture to higher pressure and temperature, thereby initiating chemical reactions promoting ignition. During these experiments, the pressure is continuously monitored and is used to deduce significant events such as the end of compression and the onset of ignition. The pressure profile is also used to assess the temperature evolution of the gas mixture with time using the adiabatic core hypothesis and the heat capacity ratio of the gas mixture. In such RCM studies, real transportation fuels containing many components are often represented by simpler surrogate fuels. While simpler surrogates such as primary reference fuels (PRFs) and ternary primary reference fuel (TPRFs) can match research and motor octane number of transportation fuels, they may not accurately replicate thermodynamic properties (including heat capacity ratio). This non-conformity could exhibit significant discrepancies in the end of compression temperature, thereby affecting ignition delay (τign) measurements. Another aspect of RCMs that can affect τign measurement is post compression heat loss, which depends on various RCM parameters including geometry, extent of insulation, pre-heating temperature etc. To, better understand the effects of these non-chemical kinetic parameters on τign, thermodynamic properties of a number of FACE G gasoline surrogates were calculated and simulated in a multi-zone RCM model. The problem was further investigated using a variance based analysis and individual sensitivities were calculated. This study highlights the effects on τign due to thermodynamic properties of

  20. Impact of thermodynamic properties and heat loss on ignition of transportation fuels in rapid compression machines

    KAUST Repository

    Ahmed, Ahfaz; Hantouche, Mireille; Khurshid, Muneeb; Mohamed, Samah; Nasir, Ehson Fawad; Farooq, Aamir; Roberts, William L.; Knio, Omar; Sarathy, Mani

    2018-01-01

    Rapid compression machines (RCM) are extensively used to study autoignition of a wide variety of fuels at engine relevant conditions. Fuels ranging from pure species to full boiling range gasoline and diesel can be studied in an RCM to develop a better understanding of autoignition kinetics in low to intermediate temperature ranges. In an RCM, autoignition is achieved by compressing a fuel/oxidizer mixture to higher pressure and temperature, thereby initiating chemical reactions promoting ignition. During these experiments, the pressure is continuously monitored and is used to deduce significant events such as the end of compression and the onset of ignition. The pressure profile is also used to assess the temperature evolution of the gas mixture with time using the adiabatic core hypothesis and the heat capacity ratio of the gas mixture. In such RCM studies, real transportation fuels containing many components are often represented by simpler surrogate fuels. While simpler surrogates such as primary reference fuels (PRFs) and ternary primary reference fuel (TPRFs) can match research and motor octane number of transportation fuels, they may not accurately replicate thermodynamic properties (including heat capacity ratio). This non-conformity could exhibit significant discrepancies in the end of compression temperature, thereby affecting ignition delay (τign) measurements. Another aspect of RCMs that can affect τign measurement is post compression heat loss, which depends on various RCM parameters including geometry, extent of insulation, pre-heating temperature etc. To, better understand the effects of these non-chemical kinetic parameters on τign, thermodynamic properties of a number of FACE G gasoline surrogates were calculated and simulated in a multi-zone RCM model. The problem was further investigated using a variance based analysis and individual sensitivities were calculated. This study highlights the effects on τign due to thermodynamic properties of

  1. Review of the KBS II plan for handling and final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    1980-01-01

    The Swedish utilities programme for disposal of spent nuclear fuel elements (KBS II) is summarized. Comments and criticism to the programme are given by experts from several foreign or international institutions. (L.E.)

  2. Universal machine ''Shtrek'' and the tractor-lifter with pneumatic-equipment control. [Auxiliary multipurpose materials handling equipment

    Energy Technology Data Exchange (ETDEWEB)

    Bal' bert, B M; Borumenskiy, V A; Lishenko, A P; Mitchenko, G A

    1982-01-01

    The machine ''Shtrek'' is described. It makes it possible to mechanize over 20 auxiliary operations: loading-unloading operations: extraction of old and deformed timbering; dissmantling of obstructions; erection of different types of timbering; making and restoring of drainage channels; laying and straightening of a drift and its leveling; assembly and disassembly of pipelines and mine equipment, etc. Depending on the type of operation, the machine has the corresponding suspended equipment. The elementary variant has a limited area of application at mines of the central region of the Dunbass. Currently a pneumatic variant of the machine ''Shtrek'' has been developed. The electric motor and the starter of the pumping equipment of the machine have been replaced by a pneumatic motor and pneumatically controlled valve KTM-50. In this case there was significant reduction in the weight of the pumping equipment and in its overall dimensions; the electric drive of the hydraulic distributors for controlling the mechanisms were replaced by simpler pneumatic ones; the logical circuit of the control system was constructed on the USEPPA elements. A specialized tractor-lifter designed for moving suspended loads is described for auxiliary operations in the near-face zone of the preparatory drifts. The machine also lifts and lowers the boom, rotates the boom by 270/sup 0/ and additionally lifts and lowers the weight-lifting hook.

  3. Concept for a LNG Gas Handling System for a Dual Fuel Engine

    Directory of Open Access Journals (Sweden)

    Michael Rachow

    2017-09-01

    Full Text Available Nowadays, ships are using LNG as main engine fuel because based on the facts that LNG has no sulphur content, and its combustion process, LNG produces low NOx content compared to heavy fuel oil and marine diesel oil. LNG is not only produces low gas emission, but may have economic advantages. In the engine laboratory of maritime studies department in Warnemunde, Germany, there is a diesel engine type MAN 6L23/30 A, where the mode operation of these engine would be changed to dual fuel engine mode operation. Therefore, in this thesis, the use dual fuel engine will be compared where it will utilize natural gas and marine diesel oil and select the required components for fuel gas supply system. By conducting the process calculation, engine MAN 6L23/30 A requires the capacity natural gas of 12.908  for 5 days at full load. A concept for LNG supply system would be arranged from storage tank until engine manifold. Germanischer Lloyd and Project Guide of dual fuel engine will be used as a guidelines to develop an optimal design and arrangement which comply with the regulation.

  4. Automatic refueling platform and CRD remote handling device for BWR plant

    International Nuclear Information System (INIS)

    Kato, Hiroaki; Takagi, Kaoru

    1978-01-01

    In BWR plants, machines for replacing fuel assemblies and control rod drives are usually operated directly by personnel. An automatic refueling platform and a CRD remote handling device aiming at radiation exposure reduction and handling perfectness are described, which are already used in BWR plants. Automation of the former is achieved in transporting fuel assemblies between a reactor pressure vessel and a fuel storage pool, shuffling fuel assemblies in a reactor core and moving fuel assemblies in a fuel storage pool. In the latter, replacement of CRDs is nearly all performed remotely. (Mori, K.)

  5. Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations

    International Nuclear Information System (INIS)

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J.

    1991-12-01

    This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 x 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990

  6. A computer model to predict temperatures and gas flows during AGR fuel handling

    International Nuclear Information System (INIS)

    Bishop, D.C.; Bowler, P.G.

    1986-01-01

    The paper describes the development of a comprehensive computer model (HOSTAGE) that has been developed for the Heysham II/Torness AGRs to predict temperature transients for all the important components during normal and fault conditions. It models not only the charge and discharge or fuel from an on-load reactor but also follows the fuel down the rest of the fuel route until it is dismantled. The main features of the physical model of gas and heat flow are described. Experimental results are used where appropriate and an indication will be given of how the predictions by HOSTAGE correlate with operating AGR reactors. The role of HOSTAGE in the Heysham II/Torness safety case is briefly discussed. (author)

  7. Transfer hood for handling fuel assemblies in nuclear reactors and especially fast reactors

    International Nuclear Information System (INIS)

    Aubert, M.; Merland, D.; Renaux, C.

    1975-01-01

    A description is given of a hood for transferring fuel assemblies between two or more separate guide ramps inclined to the vertical at the same angle of slope and extending respectively through a first passage into the reactor vessel and through a second passage into a fuel-assembly storage chamber. The hood comprises at least one shielded duct joined to a flanged rotating portion which is so arranged that the open lower end of the shielded duct is positioned in the line of extension of one guide ramp and then the other as a result of displacement of the rotating portion

  8. Work carried out in France on the design, manufacture, handling and development of nuclear fuel

    International Nuclear Information System (INIS)

    Brandt, R.C.; Joly, G.; Gloaguen, A.; Delafosse, J.

    1977-01-01

    Although the ordinary water reactors to be found in France all belong to the PWR type, the fuel used covers a broad range: box assemblies with steel canning at the SENA plant, 15x15 at TIHANGE, 17x17 for 900 MW phases, slug and plate fuel developed by the Atomic Energy Commission and extra-long 17x17 for 1300 MW phases, also being developed. A description of what France is undertaking today with respect to: 1) design; 2) manufacture; 3) management; and 4) development of full assemblies is presented [fr

  9. Demonstrating the compatibility of Canflex fuel bundles with a CANDU 6 fuelling machine

    Energy Technology Data Exchange (ETDEWEB)

    Alavi, P; Oldaker, I E [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Suk, H C; Choi, C B [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1997-12-31

    CANFLEX is a new 43-element fuel bundle, designed for high operating margins. It has many small-diameter elements in its two outer rings, and large-diameter elements in its centre rings. By this means, the linear heat ratings are lower than those of standard 37-element bundles for similar power outputs. A necessary part of the out-reactor qualification program for the CANFLEX fuel bundle design, is a demonstration of the bundle`s compatibility with the mechanical components in a CANDU 6 Fuelling Machine (FM) under typical conditions of pressure, flow and temperature. The diameter of the CANFLEX bundle is the same as that of a 37-element bundle, but the smaller-diameter elements in the outer ring result in a slightly larger end-plate diameter. Therefore, to minimize any risk of unanticipated damage to the CANDU 6 FM sidestops, a series of measurements and static laboratory tests were undertaken prior to the fuelling machine tests. The tests and measurements showed that; a) the CANFLEX bundle end plate is compatible with the FM sidestops, b) all the dimensions of the CANFLEX fuel bundle are within the specified limits. (author). 3 tabs., 3 figs.

  10. Improvements in the devices for handling the fuel of atomic reactors

    International Nuclear Information System (INIS)

    1973-01-01

    A handling device of the type associated to the platform of a travelling bridge and comprising a head which is integral with an axially fixed but angularly adjustable outer tube wherein is coaxially mounted an inner tube adapted to slide with respect to the first tube, said inner tube carrying a gripping device at the free end thereof characterized in that the inner tube is angularly fixed with respect to the outer tube, while that tube and the head taken as a whole contain a mechanism for axially displacing the inner tube, the unit formed by the tube end the head being pivotally mounted with respect to the platform [fr

  11. Storage, handling and internal transport of radioactive materials (fuel elements excepted) in nuclear power plants

    International Nuclear Information System (INIS)

    1983-06-01

    The rule applies to storage and handling as well as to transport within the plant and to the exchange of - solid radioactive wastes, - liquid radioactive wastes, except for those covered by the rule KTA 3603, - radioactive components and parts which are planned to be mounted and dismounted until shutdown of the plant, - radioactive-contaminated tools and appliances, - radioactive preparations. The rule is to be applied within the fenced-in sites of stationary nuclear power plants with LWR or HTR including their transport load halls, as fas as these are situated so as to be approachable from the nuclear power station by local transport systems. (orig./HP) [de

  12. The curium tagging approach for enhanced safeguards for spent fuel handling and reprocessing

    International Nuclear Information System (INIS)

    Menlove, H.O.; Beddingfield, D.H.; Rinard, P.M.; Wenz, T.R.

    1999-01-01

    Because of the intense neutron emission rate from curium, it can be a useful signature to measure and track special nuclear material in spent fuel and waste. By measuring the concentration of curium as well as plutonium and uranium in spent fuel streams, the ratio of curium to plutonium, uranium or other actinides can be used for waste assay. The quantity of special nuclear material in the waste streams such as the leached hulls and vitrified high-level waste can be calculated from the ratio of the curium to the plutonium, etc. The quantity of curium can be measured from the neutron emission rate from the various waste streams in the presence of the high-level gamma-ray backgrounds from fission products. Examples of using the curium ratio technique for measuring plutonium in leached hulls and vitrified waste canisters are presented in this paper. (author)

  13. Underwater fuel handling equipment maintenance. Verification of design assumptions, specific problems and tools, case study

    International Nuclear Information System (INIS)

    Kurek, J.B.

    1995-01-01

    The majority of CANDU Fuel Transfer System equipment at Pickering is located under fourteen feet of water, as dictated by the containment and shielding requirements. Such arrangement, however, creates specific problems with equipment maintenance. Each single piece of equipment serves two generating units, which means in case of defect- double losses on production, or two units shut down simultaneously for planned maintenance. The requirement for underwater maintenance was not anticipated at the design stage, which multiples the level of difficulty, and creates requirement for developing special tools for each work. Removal of the damaged fuel from the receiving bays and decontamination of submerged equipment is also part of the problem. The purpose of this presentation is to share our experience with the designers, operators, maintenance mechanics and technical personnel of the other CANDU generating stations

  14. 3rd International Conference on Stability and Handling of Liquid Fuels

    Science.gov (United States)

    1988-12-07

    potentiometric titration methods for the quantification of low levels of strong acids in gas oils are described, and results from a variety of straight-run...was homogenized and a Karl Fischer w..ater titration was performed. CONTAINING IWATEP, ’Enrecircrens w..i wh ater con tent cf greater than ’. were...storage facilities [491 M P Reynuurs, P Stegmann ........................... 137 Microbiological test methods for fuels in the laboratory and on site [5] E

  15. Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    1978-01-01

    The report gives a general summary of the Swedish KBS-project on management and disposal of vitrified reprocessed waste. Its final aim is to demostrate that the means of processing and managing power reactor waste in an absolutely safe way, as stipulated in the Swedish so called Conditions Act, already exist. Chapters on Storage facility for spent fuel, Intermidiate storage of reprocessed waste, Geology, Final repository, Transportation, Protection, and Siting. (L.E.)

  16. Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    1978-01-01

    A summary of the planning of transportation and plant design in the Swedish KBS project on management and disposal reprocessed radioactive waste. It describes a transportation system, a central storage facility for used fuel elements, a plant for intermediate storage and encapsulation and a final repository for the vitrified waste. Accounts are given for the reprocessing and vitrification. The safety of the entire system is discussed

  17. Survey and evaluation of handling and disposing of solid low-level nuclear fuel cycle wastes

    International Nuclear Information System (INIS)

    Mullarkey, T.B.; Jentz, T.L.; Connelly, J.M.; Kane, J.P.

    1976-10-01

    The report identifies the types and quantities of low-level solid radwaste for each portion of the nuclear fuel cycle, based on operating experiences at existing sites and design information for future installations. These facts are used to evaluate reference 1000 MWe reactor plants in terms of solid radwaste generation. The effect of waste volumes on disposal methods and land usage has also been determined, based on projections of nuclear power growth through the year 2000. The relative advantages of volume reduction alternatives are included. Major conclusions are drawn concerning available land burial space, light water reactors and fuel fabrication and reprocessing facilities. Study was conducted under the direction of an industry task force and the National Environmental Studies Project, a technical program of the Atomic Industrial Forum. Data was obtained from questionnaires sent to 8 fuel fabrication facilities, 39 reactor sites and 6 commercial waste disposal sites. Additional data were gathered from interviews with architect engineering firms, site visits, contacts with regulatory agencies and published literature

  18. Microcomputer simulation model for facility performance assessment: a case study of nuclear spent fuel handling facility operations

    International Nuclear Information System (INIS)

    Chockie, A.D.; Hostick, C.J.; Otis, P.T.

    1985-10-01

    A microcomputer based simulation model was recently developed at the Pacific Northwest Laboratory (PNL) to assist in the evaluation of design alternatives for a proposed facility to receive, consolidate and store nuclear spent fuel from US commercial power plants. Previous performance assessments were limited to deterministic calculations and Gantt chart representations of the facility operations. To insure that the design of the facility will be adequate to meet the specified throughput requirements, the simulation model was used to analyze such factors as material flow, equipment capability and the interface between the MRS facility and the nuclear waste transportation system. The simulation analysis model was based on commercially available software and application programs designed to represent the MRS waste handling facility operations. The results of the evaluation were used by the design review team at PNL to identify areas where design modifications should be considered. 4 figs

  19. Integration of Forest Fuel Handling in the Ordinary Forestry. Studies on Forestry, Technology and Economy of Forest Fuel Production in Lithuania

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Lars [Regional Forestry Board of Vaermland-Oerebro, Karlstad (Sweden); Budrys, Renatas [Lithuanian Forest Research Inst. (Lithuania)

    2002-07-01

    During the year 2000, The Swedish Forest Administration and Forest Department, Ministry of Environment in Lithuania, started a bilateral co-operation project, named: 'Swedish Lithuanian Wood Fuel Development Project', financed by the Swedish Energy Agency. The project was divided into 2 phases. The first phase objectives were to make a feasibility study in the eastern part of Lithuania and to identify the present conditions for the utilization of wood fuel within seven state forest enterprises and to define a demonstration and experimental area for the phase 2. The purpose of this work was to find solutions for creating horizontal and vertical integration in the handling of forest fuels in ordinary forestry and supply systems. The aim would be to give specific recommendations on which methods are the most suitable and profitable and on what type of equipment to use for various conditions and by the means of demonstrations to show how to integrate the positive results into the ordinary forestry activities. Different kinds of activities have been carried out to ensure capacity building and development on other levels within the system. 3 activity groups were established and have been working side by side with the appointed team leaders for each activity group from the institutions leading in the specific area within the forest sector in Lithuania. Swedish specialists from the Swedish Forest Administration were involved into the project and the activity groups as well. Lithuanian Forest Research Institute was involved into the project with research support. Additional to the project a mobile drum wood chipper was purchased from Sweden. 3 separate investigations have been conducted, one by Kaunas Univ. of Tech. on the analysis and estimation of material balance in Lithuania saw milling industry, another by Forest Economy Centre on wood fuel produced in industry in Lithuania and the third one by Lithuanian Energy Institute and AF international on Bio fuel

  20. Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 1. Executive summary and text

    International Nuclear Information System (INIS)

    1979-08-01

    The Generic Environmental Impact Statement on spent fuel storage was prepared by the Nuclear Regulatory Commission staff in response to a directive from the Commissioners published in the Federal Register, September 16, 1975 (40 FR 42801). The Commission directed the staff to analyze alternatives for the handling and storage of spent light water power reactor fuel with particular emphasis on developing long range policy. Accordingly, the scope of this statement examines alternative methods of spent fuel storage as well as the possible restriction or termination of the generation of spent fuel through nuclear power plant shutdown. Volume 1 includes the executive summary and the text

  1. Handling machine breakdown for dynamic scheduling by a colony of cognitive agents in a holonic manufacturing framework

    Directory of Open Access Journals (Sweden)

    T. K. Jana

    2015-09-01

    Full Text Available There is an ever increasing need of providing quick, yet improved solution to dynamic scheduling by better responsiveness following simple coordination mechanism to better adapt to the changing environments. In this endeavor, a cognitive agent based approach is proposed to deal with machine failure. A Multi Agent based Holonic Adaptive Scheduling (MAHoAS architecture is developed to frame the schedule by explicit communication between the product holons and the resource holons in association with the integrated process planning and scheduling (IPPS holon under normal situation. In the event of breakdown of a resource, the cooperation is sought by implicit communication. Inspired by the cognitive behavior of human being, a cognitive decision making scheme is proposed that reallocates the incomplete task to another resource in the most optimized manner and tries to expedite the processing in view of machine failure. A metamorphic algorithm is developed and implemented in Oracle 9i to identify the best candidate resource for task re-allocation. Integrated approach to process planning and scheduling realized under Multi Agent System (MAS framework facilitates dynamic scheduling with improved performance under such situations. The responsiveness of the resources having cognitive capabilities helps to overcome the adverse consequences of resource failure in a better way.

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

    International Nuclear Information System (INIS)

    Rajagopalan, C.; Venugopal, S.

    2013-01-01

    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

  3. Handling and storage of high-level liquid wastes from reprocessing of spent fuel

    International Nuclear Information System (INIS)

    Finsterwalder, L.

    1982-01-01

    The high level liquid wastes arise from the reprocessing of irradiated nuclear fuels, which are dissolved in aqueous acid solution, and the plutonium and unburned uranium removed in the chemical separation plant. The remaining solution, containing more than 99% of the dissolved fission products, together with impurities from cladding materials, corrosion products, traces of unseparated plutonium and uranium and most of the transuranic elements, constitutes the high-level waste. At present, these liquid wastes are usually concentrated by evaporation and stored as an aqueous nitric acid solution in high-integrity stainless-steel tanks. There is now world-wide agreement that, for the long term, these liquid wastes should be converted to solid form and much work is in progress to develop techniques for the solidification of these wastes. This paper considers the design requirements for such facilities and the experience gained during nearly 30 years of operation. (orig./RW)

  4. Transforming criticality control methods for EBR-II fuel handling during reactor decommissioning

    International Nuclear Information System (INIS)

    Eberle, C.S.; Dean, E.M.; Angelo, P.L.

    1995-01-01

    A review of the Department of Energy (DOE) request to decommission the Experimental Breeder Reactor-II (EBR-II) was conducted in order to develop a scope of work and analysis method for performing the safety review of the facility. Evaluation of the current national standards, DOE orders, EBR-II nuclear safeguards and criticality control practices showed that a decommissioning policy for maintaining criticality safety during a long term fuel transfer process did not exist. The purpose of this research was to provide a technical basis for transforming the reactor from an instrumentation and measurement controlled system to a system that provides both physical constraint and administrative controls to prevent criticality accidents. Essentially, this was done by modifying the reactor core configuration, reactor operations procedures and system instrumentation to meet the safety practices of ANS-8.1-1983. Subcritical limits were determined by applying established liquid metal reactor methods for both the experimental and computational validations

  5. Nukem Nuclear GmbH activity in CIS countries in the sphere of radioactive wastes and nuclear fuel handling

    International Nuclear Information System (INIS)

    Vaihard, A.

    1997-01-01

    NUKEM was founded in 1960 as one of the first nuclear companies in the German Federal Republic. With this work, Nukem developed not only processes for producing fuels and fuel elements, but also the plant and equipment necessary for this production. NUKEM engineers further planned and built the total infrastructure for operation of these manufacturing plants, including the supply and waste plants, the nuclear ventilation technology, the laboratory and the remote handling manipulators. The scope of activities extends from the design to the manufacture and construction of turnkey plants. The points of emphasis are plants and processes for the Treatment of radioactive wastes, storage and container technology, the decommissioning of nuclear plants, the planning and building of nuclear laboratories, the design of fuel elements and safety and monitoring technology. NUKEM Nuclear Technology is an independent division within the plant construction of the NUKEM Group. Additionally, five further subsidiary and partner companies have a spectrum of nuclear technology activities. Altogether, Nukem Nuclear Technology counts around 300 highly qualified engineers, scientists and technicians. Numerous Designs and patents underline the strength of innovative output in this area. The engineering service offered by NUKEM includes the whole spectrum of process and technology as well as construction and start-up as general engineer or general contractor: Basic engineering; Detail engineering; Procurement; Personnel Training; Start-up. Engineering and safety for nuclear technology: Process and plant planing; Media supply and disposal; Building and architecture; Electrical, measurement and control technology; Safety and accident analysis; Licensing procedures. Treatment of Radioactive Wastes: - Volume reduction of soil and liquid wastes: vaporizer plants; evaporator plants; incineration plants; pyrolysis plants; compactors. - Chemical/physical processes for residue treatment: boric acid

  6. Sustainable machining

    CERN Document Server

    2017-01-01

    This book provides an overview on current sustainable machining. Its chapters cover the concept in economic, social and environmental dimensions. It provides the reader with proper ways to handle several pollutants produced during the machining process. The book is useful on both undergraduate and postgraduate levels and it is of interest to all those working with manufacturing and machining technology.

  7. Assessment of the impact of fueling machine failure on the safety of the CANDU-PHWR

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.

    1982-01-01

    A survey of possible LOCA (Loss-of-Coolant Accident) initiating events that might take place for CANDU-PHWRs (Canadian Deuterium Uranium-Pressurized Heavy Water Reactors) has been conducted covering both direct and indirect initiators. Among the 22 initiating events that were surveyed in this study, four direct initiators have been selected and analyzed briefly. Those selected were a pump suction piping break, an isolation valve piping break, a bleed valve failure, and a fueling machine interface failure. These were selected as examples of failures that could take place in the inlet side, outlet side, or PHTS (Primary Heat Transport System) interfaces. The Pickering NGS (Unit-A) was used for this case study. Double failure (failure of the protective devices to operate when the process equipment fault occurs) and a triple failure (failure of the protective devices and the ECCS as well as the process equipment) were found to be highly improbable

  8. Superphenix 1 primary handling system fabrication and testing

    International Nuclear Information System (INIS)

    Branchu, J.; Ebbinghaus, K.; Gigarel, C.

    1985-01-01

    Primary handling covers the operations performed for spent fuel removal, new fuel insertion, and the insodium storage outside the new or spent fuel vessel. This equipment typifies many of the difficulties encountered with the project as a whole: fabrication coordination when several countries are involved and design and construction of very large, relatively complex components. Detailed design studies were mainly influenced by thermal and seismic requirements, as applicable to sodium-immersed structures. Where possible, well-tried mechanical solutions were used, but widely differing techniques were involved, ranging from the high precision fabrication of structures and mechanisms comprising numerous component parts, implying complex machining operations. No particular problems were encountered during the sodium testing of the primary handling equipment. Trends for the 1500-MW (electric) breeder include investigation of the advisability of fuel storage in the core lattice and the possibility of handling system simplification

  9. CANDU 9 fuelling machine carriage

    Energy Technology Data Exchange (ETDEWEB)

    Ullrich, D J; Slavik, J F [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1997-12-31

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs.

  10. CANDU 9 fuelling machine carriage

    International Nuclear Information System (INIS)

    Ullrich, D.J.; Slavik, J.F.

    1996-01-01

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs

  11. Seismic analysis during development stage of CANDU Model 2 fueling machine design

    International Nuclear Information System (INIS)

    Lee, L.S.S.; Mansfield, R.A.

    1989-01-01

    The CANDU Model 3 is a new small reactor presently being designed. This reactor is 450 MWe, and as with current operating CANDU's, is based on a heavy water moderated and cooled system using on-power fuelling for the once-through natural uranium fuel cycle. The CANDU 3 Standard plant is designed to be adaptable to a range of world-wide site conditions, i.e. for a peak ground acceleration of 0.3 g and a wide range of soft, medium and hard foundation medium properties. Consequently, a conservatism in the design of structure and equipment is accounted by using enveloped floor response spectra generated by the soil-structure interaction analysis. Seismic qualification of the fuelling machine (F/M) and its support structure are an essential design requirement for maintaining the integrity of the reactor coolant heat transport system (HTS) pressure boundary and the service ports penetrating the containment structure during on-power fueling. This paper deals with the initial conceptual phase of design where the details of the design are in fundamental outline form only and basic mass distribution plus layout geometry is defined

  12. Role of non-destructive examinations in leak testing of glove boxes for industrial scale plutonium handling at nuclear fuel fabrication facility along with case study

    International Nuclear Information System (INIS)

    Aher, Sachin

    2015-01-01

    Non Destructive Examinations has the prominent role at Nuclear Fuel Fabrication Facilities. Specifically NDE has contributed at utmost stratum in Leak Testing of Glove Boxes and qualifying them as a Class-I confinement for safe Plutonium handling at industrial scale. Advanced Fuel Fabrication Facility, BARC, Tarapur is engaged in fabrication of Plutonium based MOX (PuO 2 , DDUO 2 ) fuel with different enrichments for first core of PFBR reactor. Alpha- Leak Tight Glove Boxes along with HEPA Filters and dynamic ventilation form the promising engineering system for safe and reliable handling of plutonium bearing materials considering the radiotoxicity and risk associated with handling of plutonium. Leak Testing of Glove Boxes which involves the leak detection, leak rectification and leak quantifications is major challenging task. To accomplish this challenge, various Non Destructive Testing methods have assisted in promising way to achieve the stringent leak rate criterion for commissioning of Glove Box facilities for plutonium handling. This paper highlights the Role of various NDE techniques like Soap Solution Test, Argon Sniffer Test, Pressure Drop/Rise Test etc. in Glove Box Leak Testing along with procedure and methodology for effective rectification of leakage points. A Flow Chart consisting of Glove Box leak testing procedure starting from preliminary stage up to qualification stage along with a case study and observations are discussed in this paper. (author)

  13. Development of remote handling tools for glove box

    International Nuclear Information System (INIS)

    Tomita, Yutaka; Nemoto, Takeshi; Denuma, Akio; Todokoro, Akio

    1996-01-01

    For a part of advanced nuclear fuel recycling technology development, we will separate and recover Americium from the MOX fuel scrap by solvent extraction. When we carry this examination, reduction of exposure from Americium-241 is one of important problems. To solve this problem fundamentally, we studied many joints type of the remote handling tools for glove box and produced a trial production machine. Also, we carried out basic function examinations of it. As a result, we got the prospect of development of the remote handling tools which could treat Americium in glove box. (author)

  14. CanDan 2, phase 2. Final report. [Fuel cell systems for back-up power and materials handling applications]; CanDan 2, fase 2. Slutrapport

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-11-01

    CanDan 2 Phase 2 is the second phase of a research and demonstration project for fuel cell backup power systems and fuel cell powered material handling equipment. In the Backup Power segment the fuel cell units have been developed, certified and delivered. A total of 32 fuel cell backup power systems have been delivered for EnergiMidt and in operation since early 2011. Following this project EnergiMidt has purchased another 31 systems in order to make a full transition from battery backup to fuel cell backup in their entire broadband network. In the material handling segment a 10 kW fuel cell system has been fully integrated in the fork lift truck, Dantruck 3000 Power Hydrogen. The result was a much more commercially mature product than expected from the beginning of the project. The result is a finished 2,5T fork lift truck which was presented at the CE-mat fair in April 2011. (LN)

  15. Development of advanced loop-type fast reactor in Japan (4): An advanced design of the fuel handling system for the enhanced economic competitiveness

    International Nuclear Information System (INIS)

    Usui, S.; Mihara, T.; Obata, H.; Kotake, S.

    2008-01-01

    Refueling operation of sodium fast reactor (SFR) is one of major technical issue due to the chemical activities and opaqueness of sodium coolant properties in comparison with that of LWR. In the Japan Atomic Energy Agency (JAEA) sodium cooled Fast Reactor (JSFR) design study, the further reliable and rational fuel handling system (FHS) has been developing based on the experience of safe and reliable fuel handling operation in the existent SFR plants. Some of advanced concepts for the FHS have being studied in order to increase economic competitiveness further by attempting reduction of the amount of the material and the refueling time, and are scheduled to execute elemental tests and/or mock-up tests to confirm their feasibilities. (authors)

  16. Historical summary of the fuel and waste handling and disposition activities of the TMI-2 Information and Examination Program (1980-1988)

    International Nuclear Information System (INIS)

    Reno, H.W.; Schmitt, R.C.

    1988-10-01

    This report is a historical summary of the major activities conducted by the TMI-2 Information and Examination Program in managing fuel and special radioactive wastes resulting from the accident at the Unit 2 reactor of the Three Mile Island Nuclear Power Station (TMI-2). The activities often required the development and use of advanced handling, processing, and/or disposal technologies for those wastes

  17. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

    Purpose: To increase the spent fuel storage capacity and reduce the installation cost in a nuclear fuel storage facility. Constitution: Fuels handled in the nuclear fuel storage device of the present invention include the following four types: (1) fresh fuels, (2) 100 % reactor core charged fuels, (3) spent fuels just after taking out and (4) fuels after a certain period (for example one half-year) from taking out of the reactor. Reactivity is high for the fuels (1), and some of fuels (2), while low in the fuels (3) (4), Source intensity is strong for the fuels (3) and some of the fuels (2), while it is low for the fuels (1) and (4). Taking notice of the fact that the reactivity, radioactive source intensity and generated after heat are different in the respective fuels, the size of the pool and the storage capacity are increased by the divided storage control. While on the other hand, since the division is made in one identical pool, the control method becomes important, and the working range is restricted by means of a template, interlock, etc., the operation mode of the handling machine is divided into four, etc. for preventing errors. (Kamimura, M.)

  18. Robotic refueling machine

    International Nuclear Information System (INIS)

    Challberg, R.C.; Jones, C.R.

    1996-01-01

    One of the longest critical path operations performed during the outage is removing and replacing the fuel. A design is currently under development for a refueling machine which would allow faster, fully automated operation and would also allow the handling of two fuel assemblies at the same time. This design is different from current designs, (a) because of its lighter weight, making increased acceleration and speed possible, (b) because of its control system which makes locating the fuel assembly more dependable and faster, and (c) because of its dual handling system allowing simultaneous fuel movements. The new design uses two robotic arms to span a designated area of the vessel and the fuel storage area. Attached to the end of each robotic arm is a lightweight telescoping mast with a pendant attached to the end of each mast. The pendant acts as the base unit, allowing attachment of any number of end effectors depending on the servicing or inspection operation. Housed within the pendant are two television cameras used for the positioning control system. The control system is adapted from the robotics field using the technology known as machine vision, which provides both object and character recognition techniques to enable relative position control rather than absolute position control as in past designs. The pendant also contains thrusters that are used for fast, short distance, precise positioning. The new refueling machine system design is capable of a complete off load and reload of an 872 element core in about 5.3 days compared to 13 days for a conventional system

  19. Irradiation of a CANDU UOsub(2) fuel element with twenty-three machined slits cut through the zircaloy sheath

    International Nuclear Information System (INIS)

    DaSilva, R.L.

    1984-09-01

    A CANDU fuel element was purposely defected, exposing a minimum UOsub(2) fuel stack area of 272 mmsup(2), by machining 23 longitudinal slits through the Zircaloy-4 sheathing. The element was then irradiated in the X-2 loop of the NRX reactor for a period of 14.64 effective full power days at a linear heat rating of 48 kW/m to investigate the relationship between fission product release and UOsub(2) oxidation behaviour in an element with minimal fuel-to-gap fission gas trapping. The fission product releases, as measured by on-line gamma-ray spectroscopy, revealed that the noble gases and radioiodines are both released from the UOsub(2) fuel matrix directly to the coolant via simple diffusion kinetics, and that their diffusivities in hyperstoichiometric UOsub(2) are approximately equal. The oxidation of UOsub(2) to the higher states UOsub(2+x), Usub(4)Osub(9) and Usub(3)Osub(8), was accompanied by substantial fuel swelling and sheath deformation preferentially located in the lower powered end of the element. The spalling and erosion behaviour of the fuel pellets was correlated to the rate of fuel oxidation

  20. 2010 Manufacturing Readiness Assessment Update to the 2008 Report for Fuel Cell Stacks and Systems for the Backup Power and Materials Handling Equipment Markets

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, D.; Ulsh, M.

    2012-08-01

    In 2008, the National Renewable Energy Laboratory (NREL), under contract to the US Department of Energy (DOE), conducted a manufacturing readiness assessment (MRA) of fuel cell systems and fuel cell stacks for back-up power and material handling applications (MHE). To facilitate the MRA, manufacturing readiness levels (MRL) were defined that were based on the Technology Readiness Levels previously established by the US Department of Energy (DOE). NREL assessed the extensive existing hierarchy of MRLs developed by Department of Defense (DoD) and other Federal entities, and developed a MRL scale adapted to the needs of the Fuel Cell Technologies Program (FCTP) and to the status of the fuel cell industry. The MRL ranking of a fuel cell manufacturing facility increases as the manufacturing capability transitions from laboratory prototype development through Low Rate Initial Production to Full Rate Production. DOE can use MRLs to address the economic and institutional risks associated with a ramp-up in polymer electrolyte membrane (PEM) fuel cell production. In 2010, NREL updated this assessment, including additional manufacturers, an assessment of market developments since the original report, and a comparison of MRLs between 2008 and 2010.

  1. Sipping machine control system new design to perform integrity of nuclear fuel test in Cofrentes power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palomo, M., E-mail: mpalomo@iqn.upv.es [Departamento de Ingenieria Quimica y Nuclear. Universidad Politecnica de Valencia (Spain); Urrea, M., E-mail: Matias.urrea@iberdrola.es [C.N.Cofrentes - Iberdrola Generacion S.A., Cofrentes, Valencia (Spain); Curiel, M., E-mail: m.curiel@lainsa.com [LAINSA Grupo Dominguis, Valencia (Spain); Arnaldos, A., E-mail: a.arnaldos@titaniast.com [TITANIA Servicios Tecnologicos SL, Grupo Dominguis, Valencia (Spain)

    2011-07-01

    This paper we present is related to SIPPING machine control system new design to perform integrity of nuclear fuel test. This test is a non destructive technique used for evaluating the radiated nuclear fuel coating structural integrity. It is based on the radioactive emission detection of fission elements in the reactor cooling system, using the fuel inspection equipment (SIPPING). SIPPING equipment consists of one simultaneous test bell-shaped vessel of eight fuel elements, and another one for individual element test, a control workstation and some accessories (cables, thermocouples, hoses). SIPPING inspection is carried out by means of fuel element vessel. Through air injection, water flows around the element and heat evacuation is reduced, so fuel elements temperature increases. Those elements with faults shall expelled fission components dissolved in water and/or as a gas component. The project aim is the SIPPING system control design and software based on LabVIEW, for control, monitoring and documentation of the SIPPING Test. This project shall give a major functionality to the system and, at the same time, shall facilitate the user a friendlier and interactive environment allowing: to substitute the present work platform with a real-time electronic system based on cRIO and a control software ad-hoc designed for SIPPING system; to equip new system of a major redundancy for data storage, minimising loss probability of the same. (author)

  2. Sipping machine control system new design to perform integrity of nuclear fuel test in Cofrentes power plant

    Energy Technology Data Exchange (ETDEWEB)

    Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain); Palomo, M. J. [ISIRYM, Universidad Politecnica de Valencia, Camino de Vera s/n, Valencia (Spain); Urrea, M. [Iberdrola Generacion S. A., Central Nuclear Cofrentes, Carretera Almansa Requena s/n, 04662 Cofrentes, Valencia (Spain); Vaquer, J., E-mail: m.curiel@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

    2010-10-15

    This paper related to Sipping machine control system new design to perform integrity of nuclear fuel test. This test is a non destructive technique used for evaluating the radiated nuclear fuel coating structural integrity. It is based on the radioactive emission detection of fission elements in the reactor cooling system, using the fuel inspection equipment Sipping. The equipment consists of one simultaneous test bell-shaped vessel of eight fuel elements, and another one for individual element test, a control workstation and some accessories (cables, thermocouples, hoses). Sipping inspection is carried out by means of fuel element vessel. Through air injection, water flows around the element and heat evacuation is reduced, so fuel elements temperature increases. Those elements with faults shall expelled fission components dissolved in water and/or as a gas component. The project aim is the Sipping system control design and software based on LabVIEWTM, for control, monitoring and documentation of the Sipping test. This project shall give a major functionality to the system and, at the same time, shall facilitate the user a friendlier and interactive environment allowing: 1) To substitute the present work platform with a real-time electronic system based on cRIO and a control software ad-hoc designed for Sipping system. 2) To equip new system of a major redundancy for data storage, minimising loss probability of the same. (Author)

  3. Sipping machine control system new design to perform integrity of nuclear fuel test in Cofrentes power plant

    International Nuclear Information System (INIS)

    Curiel, M.; Palomo, M. J.; Urrea, M.; Vaquer, J.

    2010-10-01

    This paper related to Sipping machine control system new design to perform integrity of nuclear fuel test. This test is a non destructive technique used for evaluating the radiated nuclear fuel coating structural integrity. It is based on the radioactive emission detection of fission elements in the reactor cooling system, using the fuel inspection equipment Sipping. The equipment consists of one simultaneous test bell-shaped vessel of eight fuel elements, and another one for individual element test, a control workstation and some accessories (cables, thermocouples, hoses). Sipping inspection is carried out by means of fuel element vessel. Through air injection, water flows around the element and heat evacuation is reduced, so fuel elements temperature increases. Those elements with faults shall expelled fission components dissolved in water and/or as a gas component. The project aim is the Sipping system control design and software based on LabVIEWTM, for control, monitoring and documentation of the Sipping test. This project shall give a major functionality to the system and, at the same time, shall facilitate the user a friendlier and interactive environment allowing: 1) To substitute the present work platform with a real-time electronic system based on cRIO and a control software ad-hoc designed for Sipping system. 2) To equip new system of a major redundancy for data storage, minimising loss probability of the same. (Author)

  4. Electricity of machine tool

    International Nuclear Information System (INIS)

    Gijeon media editorial department

    1977-10-01

    This book is divided into three parts. The first part deals with electricity machine, which can taints from generator to motor, motor a power source of machine tool, electricity machine for machine tool such as switch in main circuit, automatic machine, a knife switch and pushing button, snap switch, protection device, timer, solenoid, and rectifier. The second part handles wiring diagram. This concludes basic electricity circuit of machine tool, electricity wiring diagram in your machine like milling machine, planer and grinding machine. The third part introduces fault diagnosis of machine, which gives the practical solution according to fault diagnosis and the diagnostic method with voltage and resistance measurement by tester.

  5. Fuel transporting device

    International Nuclear Information System (INIS)

    Shiratori, Hirozo.

    1979-01-01

    Purpose: In a liquid-metal cooled reactor, to reduce the waiting time of fuel handling apparatuses and shorten the fuel exchange time. Constitution: A fuel transporting machine is arranged between a reactor vessel and an out-pile storage tank, thereby dividing the transportation line of the pot for contracting fuel and transporting the same. By assuming such a construction, the flow of fuel transportation which has heretofore been carried out through fuel transportation pipes is not limited to one direction but the take-out of fuels from the reactor and the take-in thereof from the storage tank can be carried out constantly, and much time is not required for fuel exchange. (Kamimura, M.)

  6. 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)

    Raymond, R.E.

    2011-01-01

    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

  7. A review and analysis of European industrial experience in handling LWR [light water reactor] spent fuel and vitrified high-level waste

    International Nuclear Information System (INIS)

    Blomeke, J.O.

    1988-06-01

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performances of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States. 79 refs., 71 figs., 10 tabs

  8. Assessment of pollutant emissions with GTL as a drop in fuel for medium and heavy-duty vehicles, inland shipping and non-road machines - Final

    NARCIS (Netherlands)

    Verbeek, R.P.

    2014-01-01

    Effect van Shell GTL Fuel op de uitstoot van vrachtauto’s, binnenvaartschepen en mobiele machines Paraffinische dieselbrandstoffen als GTL Fuel staan bekend om hun gunstige invloed op de uitstoot van schadelijke emissies zoals NOx en fijnstof. TNO heeft in opdracht van Shell een groot aantal

  9. Time and dose assessment of barge shipment and at-reactor handling of a CASTOR V/21 spent fuel storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Hostick, C.J. (Pacific Northwest Lab., Richland, WA (United States)); Lavender, J.C. (Westinghouse Hanford Co., Richland, WA (United States)); Wakeman, B.H. (Virginia Electric and Power Co., Richmond, VA (United States))

    1992-04-01

    This report contains the results of a time/motion analysis and a radiation dose assessment made during the receipt from barge transport and the loading of CAst iron cask for Storage and Transport Of Radioactive material (CASTOR) V/21 storage casks with spent nuclear fuel at the Surry Power Station in Virginia during 1987. The study was a cooperative effort between Pacific Northwest Laboratory (PNL) and Virginia Electric and Power Company (Virginia Power), and was funded by the US Department of Energy (DOE) Transportation Program Office. In this study, cask handling activities were tracked at the Surry Power Station, tracing the transfer of the empty spent fuel storage cask from an ocean-going vessel to a barge for river transport through the activities required to place the loaded storage cask at an at-reactor storage location.

  10. Generic environmental impact statement on handling and storage of spent light water power reactor fuel. Executive summary and text

    International Nuclear Information System (INIS)

    1978-03-01

    The study covers the following: (1) The magnitude of the possible shortage of spent fuel storage capacity; (2) The options for dealing with the problem; (3) A cost-benefit analysis of the alternatives; (4) The impacts of possible additional transportation of spent fuel that may be required should one or more of the options be adopted; (5) The need for more definitive regulations and guidance covering the licensing of one or more of the options for dealing with the problem; and (6) The possible need for amendments to 10 CFR 51.20(e). The scope of this study is limited to considerations pertinent to the interim storage of spent fuel

  11. Determination of Vaporization Properties and Volatile Hazardous Components Relevant to Kukersite Oil Shale Derived Fuel Oil Handling

    Directory of Open Access Journals (Sweden)

    Ada TRAUMANN

    2014-09-01

    Full Text Available The aim of this study was to investigate vaporization properties of shale fuel oil in relation to inhalation exposure. The shale fuel oil was obtained from kukersite oil shale. The shale oil and its light fraction (5 % of the total fuel oil were characterized by vapor pressure curve, molecular weight distribution, elemental composition and functional groups based on FTIR spectra. The rate of vaporization from the total fuel oil at different temperatures was monitored as a function of time using thermogravimetric analysis (TGA. It is shown that despite its relatively low vapor pressure at room temperature a remarkable amount of oil vaporizes influencing air quality significantly. From the TGA data the changes in the vapor pressure during vaporization process were estimated. Although the shale fuel oil has a strong, unpleasant smell, the main hazards to workplace air quality depend on the vaporization rate of different toxic compounds, such as benzene, toluene, xylene or phenolic compounds. The presence of these hazardous substances in the vapor phase of shale fuel oil was monitored using headspace analysis coupled with selective ion monitoring (SIM and confirmed by the NIST Mass Spectral library and retention times of standards. DOI: http://dx.doi.org/10.5755/j01.ms.20.3.4549

  12. Pie technique of LWR fuel cladding fracture toughness test

    International Nuclear Information System (INIS)

    Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu

    2006-01-01

    Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries Ltd. (NFI) for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine (EDM), pre-cracking by fatigue tester, sample assembling to the compact tension (CT) shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination (PIE) using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility (WASTEF). (author)

  13. Shear machines

    International Nuclear Information System (INIS)

    Astill, M.; Sunderland, A.; Waine, M.G.

    1980-01-01

    A shear machine for irradiated nuclear fuel elements has a replaceable shear assembly comprising a fuel element support block, a shear blade support and a clamp assembly which hold the fuel element to be sheared in contact with the support block. A first clamp member contacts the fuel element remote from the shear blade and a second clamp member contacts the fuel element adjacent the shear blade and is advanced towards the support block during shearing to compensate for any compression of the fuel element caused by the shear blade (U.K.)

  14. Analysed a defective of the machine for a cap-tube nuclear fuel element ME-27 from its electricity point of view

    International Nuclear Information System (INIS)

    Achmad Suntoro

    2009-01-01

    It has been analysed a defective of the machine for a cap-tube nuclear fuel element ME-27 from its electricity point of view. The machine uses magnetic force resistance welding technique. A short circuit was happened within the machine because the nut for tightening high voltage cable for welding transformer was broken so that the cable touched the machine body and produced the short circuit. This condition made both the primary circuit breaker in the building down and produced high voltage pulse induction to the electronic circuit within the machine so that one of its electronic components was defective. This case becomes warnings on how important of tightening a nut according to its strength specification (using wrench torque) and the necessity of voltage transient limitation circuit to be installed. Both of the warnings are necessary for any equipment consuming high electric current oriented such as the ME-27 machine. (author)

  15. Consolidated fuel-reprocessing program:: man/machine interface development for the REMOTEX concept

    International Nuclear Information System (INIS)

    Garin, J.; Clarke, M.M.

    1981-01-01

    This paper describes ongoing research at ORNL to develop a man/machine interface system that can be used to remotely control a system composed of a transporter base and a force-reflecting, servo-controlled manipulator. A unique feature of the concept is the incorporation of totally remote operation. Thus, a major objective is the requirement that an operator have a sense of presence in the remote environment. Man/machine interface requirements for this totally remote operation remain to be developed. Therefore, a simulator is being built to optimize such requirements and the developments are discussed

  16. Stress analysis of biomass fuel molding machine piston type stamping forming cone

    Directory of Open Access Journals (Sweden)

    Wu Gaofeng

    2015-01-01

    Full Text Available It is established the ram biomass straw machine as the analysis object in this paper,the molding machine cones of stress in the forming process of the analysis of the system. We used pottery instead of Wear-resistant cast iron for improving the performance of forming sleeve. The structure of the forming sleeve was analyzed with the mechanical module of a soft named Pro/engineer in this paper. The result indicated that the program was feasible. With the sensitivity analysis we identified the suitable angle for the sleeve.

  17. Safety assessment document for spent fuel handling, packaging, and storage demonstrations at the E-MAD facility on the Nevada Test Site

    International Nuclear Information System (INIS)

    1985-04-01

    The objectives for spent fuel handling and packaging demonstration are to develop the capability to satisfactorily encapsulate typical commercial nuclear reactor spent fuel assemblies and to establish the suitability of interim dry surface and near surface storage concepts. To accomplish these objectives, spent fuel assemblies from a pressurized water reactor have been received, encapsulated in steel canisters, and emplaced in on-site storage facilities and subjected to other tests. As an essential element of these demonstrations, a thorough safety assessment of the demonstration activities conducted at the E-MAD facility has been completed. This document describes the site location and characteristics, the existing E-MAD facility, and the facility modifications and equipment additions made specifically for the demonstrations. The document also summarizes the Quality Assurance Program utilized, and specifies the principal design criteria applicable to the facility modifications, equipment additions, and process operations. Evaluations have been made of the radiological impacts of normal operations, abnormal operations, and postulated accidents. Analyses have been performed to determine the affects on nuclear criticality safety of postulated accidents and credible natural phenomena. The consequences of postulated accidents resulting in fission product gas release have also been estimated. This document identifies the engineered safety features, procedures, and site characteristics that (1) prevent the occurrence of potential accidents or (2) assure that the consequences of postulated accidents are either insignificant or adequately mitigated

  18. Technical exercise and demonstration of the spent fuel attribute tester at the TVO NPS in Finland

    International Nuclear Information System (INIS)

    Tikkinen, J.; Tarvainen, M.

    1991-01-01

    A piece of new safeguards equipment, the Spent Fuel Attribute Tester (SFAT), is being developed for the verification of spent nuclear fuel in a standard storage geometry. Lifting of fuel assemblies from the storage position is not required for the verification. The SFAT can be handled like a fresh fuel assembly in the storage basin by the fuel handling machine. The feasibility of the SFAT-equipment for the verification of spent BWR fuel was demonstrated. A comparison of various types of gamma detectors, such as the Geiger-Mueller counter, NaI- and CdTe detectors was made for SFAT use. Measurements for optimizing the lead shielding, filtering, collimation and other geometrical parameters of SFAT were made. The precision of movements of the SFAT in the pond by the fuel handling machine and safety margins for these operations were estimated. (orig.)

  19. The Back-End of the Nuclear Fuel Cycle in Sweden. Considerations for safeguards and data handling

    International Nuclear Information System (INIS)

    Fritzell, Anni

    2011-01-01

    All nuclear facilities and activities in Sweden are under safeguards - an international monitoring system for all nuclear material. When the planned facilities for encapsulation and final disposal of spent nuclear fuel are constructed, they will also be covered by the safeguards system. The Swedish plans for final disposal is to emplace all spent fuel in a geological repository. The new facility type, the geological repository, will mean that the safeguards system is faced with new challenges, mainly since the nuclear material will be inaccessible after encapsulation and emplacement. This implies that, unlike for existing facilities, it is not possible to verify that the nuclear material is where it is declared to be or that it has the declared characteristics. This report consists of three parts, where each part investigates one aspect of safeguards for encapsulation and final disposal of spent nuclear fuel. The first part, Paper 1, presents a plausible safeguards approach for the two new facilities. The paper starts with an introduction to international safeguards and to the facilities. The facility layouts and processes are comprehensively described. The main part of Paper 1 is spent describing a safeguards system that covers all diversion paths for fissile material. The diversion paths are identified in the diversion path analysis which is the basis for Paper 3. A strategy to detect diversion is presented for each diversion path. The safeguards system comprises three main measures: 1. Verification of Nuclear Material Accountancy using, for example, verifying measurements and comparisons between shipment documents and receipt documents for transports. 2. Containment and Surveillance which are methods used to maintain continuity of knowledge of the nuclear material during periods between inspections. 3. Design Information Verification which is methods to verify that nuclear facilities are designed and operated according to declarations. The second part of the

  20. The Back-End of the Nuclear Fuel Cycle in Sweden. Considerations for safeguards and data handling

    Energy Technology Data Exchange (ETDEWEB)

    Fritzell, Anni (ES-konsult, Solna (Sweden))

    2011-01-15

    All nuclear facilities and activities in Sweden are under safeguards - an international monitoring system for all nuclear material. When the planned facilities for encapsulation and final disposal of spent nuclear fuel are constructed, they will also be covered by the safeguards system. The Swedish plans for final disposal is to emplace all spent fuel in a geological repository. The new facility type, the geological repository, will mean that the safeguards system is faced with new challenges, mainly since the nuclear material will be inaccessible after encapsulation and emplacement. This implies that, unlike for existing facilities, it is not possible to verify that the nuclear material is where it is declared to be or that it has the declared characteristics. This report consists of three parts, where each part investigates one aspect of safeguards for encapsulation and final disposal of spent nuclear fuel. The first part, Paper 1, presents a plausible safeguards approach for the two new facilities. The paper starts with an introduction to international safeguards and to the facilities. The facility layouts and processes are comprehensively described. The main part of Paper 1 is spent describing a safeguards system that covers all diversion paths for fissile material. The diversion paths are identified in the diversion path analysis which is the basis for Paper 3. A strategy to detect diversion is presented for each diversion path. The safeguards system comprises three main measures: 1. Verification of Nuclear Material Accountancy using, for example, verifying measurements and comparisons between shipment documents and receipt documents for transports. 2. Containment and Surveillance which are methods used to maintain continuity of knowledge of the nuclear material during periods between inspections. 3. Design Information Verification which is methods to verify that nuclear facilities are designed and operated according to declarations. The second part of the

  1. Development on the ex-vessel transfer machine for Monju, (1)

    International Nuclear Information System (INIS)

    Inoue, Takashi; Kinuta, Kenji; Tomita, Takaaki

    1980-01-01

    As for the fast breeder reactors being developed as the national project, the construction of the prototype reactor ''Monju'' is started in 1980 based on the results of development of the experimental reactor ''Joyo''. In the fuel-handling facility, the development of which is promoted by Fuji Electric Co., Ltd., the development of the fuel transfer machines which take out and charge the fuel for the reactors and carry out the transport between fuel storage facilities is important subject. In this paper, the outline of the development of a fuel transfer machine is described, centering around the handling of fuel in sodium and the removal of the heat of spent fuel. ''Monju'' with 300 MW electric output is being developed for the purposes of putting fast breeder reactors in practical use, increasing plant power output, improving the rate of operation, and securing the safety. Plutonium is used as the fuel, and liquid metallic sodium is used as the coolant in fast breeder reactors, accordingly the fuel must be handled safely in gas-tight, shielded vessels while it is cooled, and securely in sodium and sodium cover gas by remote operation. Gripper, gripper-driving mechanism, coffin, movable block, door valves, cooling system and carriage compose the fuel transfer machine, and these are described. The main results of development and the tests for the development, such as the trial manufacture of gripper and decay heat removal test, are reported. (Kako, I.)

  2. A prototype machine using thermal type Stirling solar energy and bio fuel as a primary energy source

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Carlos Cesar; Sousa, Regina Celia de; Santos, Jose Maria Ramos dos; Oliveira, Antonio Jose Silva [Universidade Federal do Maranhao (UFMA), Sao Luis, MA (Brazil). Dept. de Fisica

    2011-07-01

    Full text. Depending on the energy crisis and global warming became necessary to seek new sources of energy that could minimize the serious problems arising from this situation. The energy base that supported our growth in recent decades has supported - heavily on fossil fuel, highly polluting since its extraction and consumption, causing great environmental impact. Before his coal, also harmful to human health and nature. Modern life has been moved at the expense of exhaustible resources that took millions of years to form and will end one day. In this work we developed a prototype that uses a heat engine cycle of the Stirling engine with a heat source, arising from the burning of bio fuels or solar power. The main bio fuel used was ethanol. Ethanol is a product of today's diverse market applications, widely used as automotive fuel in hydrated form or blended with gasoline. The main layout of our prototype are: the four-cylinder, two for expansion and the other two for compression, a heat spreader and heat sinks. These simple components can be arranged in various configurations allowing a large space to the adequacy and efficiency of the machine. In experimental measurements made in our prototype, we have an angular speed of 360.1 rpm (revolutions per minute) with an average temperature of 215.6 deg C camera hot (expansion cylinder) and 30 deg C cold source (compression cylinders) and torque generated by our machine is 0.388 Nm Our device is multi-fuel and can be used virtually any source of energy: gasoline, ethanol, methanol, natural gas, diesel, biogas, LPG and solar energy. The construction of this device allowed us to investigate the processes of transformation of energy: chemical, thermal, and mechanical and maximize efficiency of the Stirling engine. To complete the monitoring apparatus, use equipment such as notebook, digital tachometer and a data acquisition Agilent 34970A model. These devices were used in monitoring the angular velocity and

  3. Identifying subassemblies by ultrasound to prevent fuel handling error in sodium fast reactors: First test performed in water

    International Nuclear Information System (INIS)

    Paumel, Kevin; Lhuillier, Christian

    2015-01-01

    Identifying subassemblies by ultrasound is a method that is being considered to prevent handling errors in sodium fast reactors. It is based on the reading of a code (aligned notches) engraved on the subassembly head by an emitting/receiving ultrasonic sensor. This reading is carried out in sodium with high temperature transducers. The resulting one-dimensional C-scan can be likened to a binary code expressing the subassembly type and number. The first test performed in water investigated two parameters: width and depth of the notches. The code remained legible for notches as thin as 1.6 mm wide. The impact of the depth seems minor in the range under investigation. (authors)

  4. Use of fission track analysis technique for the determination of MicroBequerel level of {sup 239}Pu in urine samples from radiation workers handling MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, J.R., E-mail: yadav_jogendra@rediffmail.co [Health Physics Laboratory, Health Physics Division, BARC, Tarapur 401502 (India); Rao, D.D.; Kumar, Ranjeet [Health Physics Laboratory, Health Physics Division, BARC, Tarapur 401502 (India); Aggarwal, S.K. [Fuel chemistry Division, BARC, Trombay, Mumbai 400085 (India)

    2011-07-15

    Fission track analysis (FTA) technique for the determination of {sup 239}Pu excreted through urine has been standardized using blank samples, tracer and {sup 239}Pu spikes. Double stage anion exchange separation protocol has been applied and an average radiochemical recovery of {sup 239}Pu of 18% was obtained. An average track registration efficiency of 11 tracks per {mu}Bq of {sup 239}Pu, irradiated to 0.35x10{sup 17} neutron fluence was established. Reagent blank urine samples from 11 controlled subjects were analyzed by FTA and an average of 149{+-}14 tracks was obtained. Minimum detectable activity of 34 {mu}Bq L{sup -1} of urine sample was obtained and will be useful for monitoring chronic exposure cases handling MOX fuel.

  5. A study on determination methods of fueling machine heavy water supply setpressure for Wolsong nuclear power plant unit 1

    International Nuclear Information System (INIS)

    Kim, J. M.; Jeong, B. Y.; Baek, S. J.; Noh, T. S.; Kim, Y. H.; Park, W. K.

    2001-01-01

    The present Wolsong 1 Fuel Handling (F/H)D 2 O Supply Pressure Control System, based on an analog cascaded Proportional-Integral-Differential (PID) control, is less accurate and requires more labor for test and maintenance in comparison with up-to-data digital controllers. Furthermore, F/H operator and technical staff have recently encountered difficulties in operation and maintenance because of frequent occurrences of system instability and failure, and obsolescence of hardware. However the analysis and design review of F/H D 2 O Supply Pressure Control System have not been performed appropriately. Therefore, the design review of F/H D 2 O Supply Pressure Control System has been thoroughly reviewed and analyzed. Based on the analysis results, the optimum pressure setpoints and its determination methods have been proposed for Wolsong Nuclear Power Plant Unit 1

  6. Development of the scientific concept of the phosphate methods for actinide-containing waste handling (pyrochemical fuel reprocessing)

    International Nuclear Information System (INIS)

    Orlova, A.I.; Orlova, V.A.; Skiba, O.V.; Bychkov, A.V.; Volkov, Yu.F.; Lukinykh, A.N.; Tomilin, S.V.; Lizin, A.A.

    2008-01-01

    Full text of publication follows: The crystallochemical phosphate concept in question is developed successfully in the new pyro-electrochemical reprocessing technology of irradiated fuel in molten chlorides of alkaline elements at one of the leading scientific nuclear centers - Research Institute of Atomic Reactors. Irradiated fuel is dissolved in molten chlorides of alkaline elements by mean of treating by chlorine. Then uranium and plutonium dioxides are removed electrochemically. The melt, when used many times, is contaminated by the residual actinide and contains fission products and the so called 'process' elements. This melt is unacceptable for future use. Phosphate methods can be applied for the solution of the following tasks: a) reprocessing (purification) of molten chloride salt solvents; b) conversion of the spent chloride melts to the insoluble stable crystalline product for safe storage and disposal. Within the framework of task 'a' phosphate methods may be realized by the several ways: 1) phosphate concentrating of impurities and their extraction from molten chlorides into solid phase by mean of chemical precipitation, co-precipitation, ion exchange and other chemical interactions, 2) conversion of precipitated waste phosphates to stable crystalline phosphate powders or ceramics for safe storage and disposal. (authors)

  7. Alloy 33: A new material for the handling of HNO3/HF media in reprocessing of nuclear fuel

    International Nuclear Information System (INIS)

    Koehler, M.; Heubner, U.; Eichenhofer, K.W.; Renner, M.

    1997-01-01

    Alloy 33, an austenitic 33Cr-32Fe-31Ni-1.6Mo-0.6Cu-0.4N material shows excellent resistance to corrosion when exposed to highly oxidizing media as e.g. HNO 3 and HNO 3 /HF mixtures which are encountered in reprocessing of nuclear fuel. According to the test results available so far, resistance to corrosion in boiling azeotropic (67%) HNO 3 is about 6 and 2 times superior to AISI 304 L and 310 L. In higher concentrated nitric acid it can be considered corrosion resistant up to 95% HNO 3 at 25 C, up to 90% HNO 3 at 50 C and up to somewhat less than 85% HNO 3 at 75 C. In 20% HNO 3 /7% HF at 50 C its resistance to corrosion is superior to AISI 316 Ti and Alloy 28 by factors of about 200 and 2.4. Other media tested with different results include 12% HNO 3 with up to 3.5% HF and 0.4% HF with 32 to 67.5% HNO 3 at 90 C. Alloy 33 is easily fabricated into all product forms required for chemical plants (e.g. plate, sheet, strip, wire, tube and flanges). Components such as dished ends and tube to tube sheet weldments have been successfully fabricated facilitating the use of Alloy 33 for reprocessing of nuclear fuel

  8. Handling wood shavings

    Energy Technology Data Exchange (ETDEWEB)

    1974-09-18

    Details of bulk handling equipment suitable for collection and compressing wood waste from commercial joinery works are discussed. The Redler Bin Discharger ensures free flow of chips from storage silo discharge prior to compression into briquettes for use as fuel or processing into chipboard.

  9. Examination process of a nuclear reactor fuel assembly and examination machine to bring the process into operation

    International Nuclear Information System (INIS)

    Delaroche, P.; Leseur, A.; Saglio, R.; Vaubert, Y.

    1983-01-01

    The machine to examine a fuel assembly of a nuclear reactor includes a support on which the assembly to be examined is placed, a source emitting waves, directed to the assembly to be examined, devices to examine the assembly to be examined that receive the emitted wave by the said source and that have been reflected by the assembly. The examination devices have an axis, this axis being directed to a mirror, this mirror being inclined in such a way that it reflects the waves reflected by the assembly to the examination devices, a radiation protection, to avoid the radiation emitted by the assembly, being diposed between the assembly and the examination devices [fr

  10. Machine to harvest slash, brush, and thinnings for fuel and fiber- a concept

    Science.gov (United States)

    P. Koch; D.W. McKenzie

    1976-01-01

    It is proposed that biomass from noncommercial thinnings, and from logging slash residual after harvest, be hogged and recovered for fuel and fiber. Such a procedure might yield two dividends of biomass totalling as much as 90,000 kg per ha (40 tons per acre, green weight basis) from each rotation of southern pine. For sites deficient in organic matter, it is...

  11. Charging machine

    International Nuclear Information System (INIS)

    Medlin, J.B.

    1976-01-01

    A charging machine for loading fuel slugs into the process tubes of a nuclear reactor includes a tubular housing connected to the process tube, a charging trough connected to the other end of the tubular housing, a device for loading the charging trough with a group of fuel slugs, means for equalizing the coolant pressure in the charging trough with the pressure in the process tubes, means for pushing the group of fuel slugs into the process tube and a latch and a seal engaging the last object in the group of fuel slugs to prevent the fuel slugs from being ejected from the process tube when the pusher is removed and to prevent pressure liquid from entering the charging machine. 3 claims, 11 drawing figures

  12. HyLIFT-FLEX. ''Development and demonstration of flexible and scalable fuel cell power system for various material handling vehicles''. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-10-15

    The project has successfully developed and tested a new fuel cell system from H2 Logic in a tow tractor from MULAG. Based on the project results a positive decision has been taken on continuing commercialisation efforts. Next step will be a large scale demonstration of up to 100 units in a new project named HyLIFT-Europe that is expected to commence in early 2013, with support from the FCH-JU programme. Main efforts in the project have been the development of a new fuel cell system, named H2Drive from H2 Logic, and the integration and test in a standard battery powered COMET 3 towing tractor from MULAG. The system size is exactly the same as a standard battery box (DIN measures) and can be easily integrated into e.g. the MULAG vehicle or other electric powered material handling vehicles using the same battery size. Several R and D efforts on the fuel cell system have been conducted with the aim to reduce cost and improve efficiency, among others the following: 1) New air compressor sub-system and control - improving overall system efficiency with {approx}2,5%. 2) New simplified air-based compressor cooling sub-system. 3) New hydrogen compressor sub-system with improved efficiency and reduced cost. 4) New hydrogen inlet and outlet manifold sub-system - resulting in reduction of more than 50% of all sensor components in the fuel cell system. 5) New DC/DC converter with an average efficiency of 97% - a 3% improvement. 6) A new optimized hybrid system that meets the vehicle cycle requirements. In total the R and D efforts have improved the overall fuel cell system efficiency with 10% and helped to reduce costs with 33% compared to the previous generation. A first prototype of the developed H2Drive system has been constructed and integrated into the MULAG Towing Tractor. Only few modifications were made on the base vehicle, among others integration of cabin-heating, displays and motor control. Several internal tests were conducted at H2 Logic and MULAG before making a

  13. Simulation and cost analysis of systems for handling of fuel straw - applied to a heating plant in Skaane; Simulering och kostnadsanalys av hanteringssystem foer braenslehalm - tillaempning foer en vaermeanlaeggning i Skaane

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Daniel

    2010-05-15

    The following conclusions were drawn from this study: - The total costs for delivery of fuel straw to heating plants ranged from about 130 SEK/MWh to 180 SEK/MWh (1 SEK approx 0.14 USD), depending on the payment to farmers for the straw, the storage method used, the profit margins of the contractors, etc. - The area investigated, which covered parts of the municipalities of Svaloev, Esloev and Kaevlinge, has very good conditions for harvest of straw compared with other parts of Sweden. The yield of winter wheat (and straw) is high, the road network is well developed, the fields are concentrated and large, the straw crops ripen early, and security of supply between years is high. - For a harvest and handling system with an average capacity of about 8 000 tonnes of straw per year, the simulations showed that a machine system with six balers, two loaders in the field, four transporters (tractor), one loader at the stores and lorry transport to the heating plant was cost-effective with regard to system performance (for example, expressed as the average amount of straw baled per year with a moisture content below 18%, the average quantity of straw per year put into storage prior to 1 October, the average lying time for the straw between threshing and baling, etc.). - Harvesting straw from various crops (in this case winter rape, rye and winter wheat) instead of only winter wheat had no major impact on costs, but increased the chance to harvest 'greyer' straw with lower contents of alkali metals, chlorine, etc. Only including fields of at least 5 ha had a minor impact on costs. Noticeable cost savings occurred when the field size limit was set to 10 ha or more. Increasing the moisture content limit at baling to 20% made it possible to harvest more straw, especially in years with difficult weather conditions. - The storage method used had a decisive impact on costs. Outdoor storage decreased the total costs to 133 SEK/MWh (7% storage losses), compared with 154

  14. Guide to improving the performance of a manipulator system for nuclear fuel handling through computer controls. Final report

    International Nuclear Information System (INIS)

    Evans, J.M. Jr.; Albus, J.S.; Barbera, A.J.; Rosenthal, R.; Truitt, W.B.

    1975-11-01

    The Office of Developmental Automation and Control Technology of the Institute for Computer Sciences and Technology of the National Bureau of Standards provides advising services, standards and guidelines on interface and computer control systems, and performance specifications for the procurement and use of computer controlled manipulators and other computer based automation systems. These outputs help other agencies and industry apply this technology to increase productivity and improve work quality by removing men from hazardous environments. In FY 74 personnel from the Oak Ridge National Laboratory visited NBS to discuss the feasibility of using computer control techniques to improve the operation of remote control manipulators in nuclear fuel reprocessing. Subsequent discussions led to an agreement for NBS to develop a conceptual design for such a computer control system for the PaR Model 3000 manipulator in the Thorium Uranium Recycle Facility (TURF) at ORNL. This report provides the required analysis and conceptual design. Complete computer programs are included for testing of computer interfaces and for actual robot control in both point-to-point and continuous path modes

  15. Comments on Frequency Swept Rotating Input Perturbation Techniques and Identification of the Fluid Force Models in Rotor/bearing/seal Systems and Fluid Handling Machines

    Science.gov (United States)

    Muszynska, Agnes; Bently, Donald E.

    1991-01-01

    Perturbation techniques used for identification of rotating system dynamic characteristics are described. A comparison between two periodic frequency-swept perturbation methods applied in identification of fluid forces of rotating machines is presented. The description of the fluid force model identified by inputting circular periodic frequency-swept force is given. This model is based on the existence and strength of the circumferential flow, most often generated by the shaft rotation. The application of the fluid force model in rotor dynamic analysis is presented. It is shown that the rotor stability is an entire rotating system property. Some areas for further research are discussed.

  16. Development and operating performance of the refuelling machine of the Fugen

    International Nuclear Information System (INIS)

    Kaneko, Jun; Kasai, Yoshimitsu; Takeshita, Norito; Ohta, Takeo

    1985-01-01

    In the advanced thermal reactor ''Fugen'' power station, with the refuelling machine the fuel replacement during operation is made through the reactor bottom. Its design was started in 1967 and up to 1975 various tests were conducted. Fugen's refuelling machine has thus been used from the initial fuel loading in 1978 and handled so far about 1300 fuel assemblies in seven times of the refuelling. In the stage of Fugen operation there occurred failure of the grab drive due to crud, etc. At present, with such troubles all eliminated, the refuelling machine is in steady operation with proper maintenance. The results with Fugen's refuelling machine are reflected in the development of the refuelling machine for the demonstration ATR. (Mori, K.)

  17. Field handling conditions of raw milk sold in vending machines: experimental evaluation of the behaviour of Listeria monocytogenes, Escherichia coli O157:H7, Salmonella Typhimurium and Campylobacter jejuni

    Directory of Open Access Journals (Sweden)

    Renato G. Zanoni

    2012-01-01

    Full Text Available The direct sale by farmers of raw milk for human consumption has been allowed in Italy since 2004. The aim of this study was to evaluate the behaviour of selected foodborne pathogens in raw milk sold in vending machines, in field handling conditions, and during shelf-life from production to consumption. Temperature of storage of raw milk in 33 farms authorized to produce and sell raw milk were investigated from farm to vending machine delivery, together with consumer habits in one province of the Emilia-Romagna region of northern Italy. Failure to maintain appropriate low temperatures during shelf-life was recorded and 43% of consumers did not boil milk before consumption. Listeria monocytogenes, Escherichia coli O157:H7, Salmonella Typhimurium and Campylobacter jejuni strains were inoculated into raw milk samples, and the best (4°C as established by law and worst temperature storage conditions detected (variable temperature were simulated. Boiling tests were performed for each pathogen considered at high and low levels of contamination. Results showed an increase in L. monocytogenes in milk stored at 4°C and at variable temperatures recorded in shelf-life monitoring, an increase in E. coli O157:H7 and S. Typhimurium at variable temperatures but not at 4°C, and a decrease in C. jejuni in all storage conditions. Boiling milk is effective in making it safe for consumers. This study provides evidence that appropriate handling of raw milk, maintaining low temperatures, together with consumer education concerning boiling raw milk before consumption are key factors in preventing foodborne infections linked to raw milk consumption, and helps assess the risk of foodborne infection linked to raw milk consumption.

  18. Waste Handling Building Conceptual Study

    International Nuclear Information System (INIS)

    G.W. Rowe

    2000-01-01

    The objective of the ''Waste Handling Building Conceptual Study'' is to develop proposed design requirements for the repository Waste Handling System in sufficient detail to allow the surface facility design to proceed to the License Application effort if the proposed requirements are approved by DOE. Proposed requirements were developed to further refine waste handling facility performance characteristics and design constraints with an emphasis on supporting modular construction, minimizing fuel inventory, and optimizing facility maintainability and dry handling operations. To meet this objective, this study attempts to provide an alternative design to the Site Recommendation design that is flexible, simple, reliable, and can be constructed in phases. The design concept will be input to the ''Modular Design/Construction and Operation Options Report'', which will address the overall program objectives and direction, including options and issues associated with transportation, the subsurface facility, and Total System Life Cycle Cost. This study (herein) is limited to the Waste Handling System and associated fuel staging system

  19. Handling of multiassembly sealed baskets between reactor storage and a remote handling facility

    International Nuclear Information System (INIS)

    Massey, J.V.; Kessler, J.H.; McSherry, A.J.

    1989-06-01

    The storage of multiple fuel assemblies in sealed (welded) dry storage baskets is gaining increasing use to augment at-reactor fuel storage capacity. Since this increasing use will place a significant number of such baskets on reactor sites, some initial downstream planning for their future handling scenarios for retrieving multi-assembly sealed baskets (MSBs) from onsite storage and transferring and shipping the fuel (and/or the baskets) to a federally operated remote handling facility (RHF). Numerous options or at-reactor and away-from-reactor handling were investigated. Materials handling flowsheets were developed along with conceptual designs for the equipment and tools required to handle and open the MSBs. The handling options were evaluated and compared to a reference case, fuel handling sequence (i.e., fuel assemblies are taken from the fuel pool, shipped to a receiving and handling facility and placed into interim storage). The main parameters analyzed are throughout, radiation dose burden and cost. In addition to evaluating the handling of MSBs, this work also evaluated handling consolidated fuel canisters (CFCs). In summary, the handling of MSBs and CFCs in the store, ship and bury fuel cycle was found to be feasible and, under some conditions, to offer significant benefits in terms of throughput, cost and safety. 14 refs., 20 figs., 24 tabs

  20. Uranium hexafluoride handling

    International Nuclear Information System (INIS)

    1991-01-01

    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 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

  1. 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.

  2. Remote handling needs of the Princeton Plasma Physics Laboratory

    International Nuclear Information System (INIS)

    Smiltnieks, V.

    1982-07-01

    This report is the result of a Task Force study commissioned by the Canadian Fusion Fuels Technology Project (CFFTP) to investigate the remote handling requirements at the Princeton Plasma Physics Laboratory (PPPL) and identify specific areas where CFFTP could offer a contractual or collaborative participation, drawing on the Canadian industrial expertise in remote handling technology. The Task Force reviewed four areas related to remote handling requirements; the TFTR facility as a whole, the service equipment required for remote maintenance, the more complex in-vessel components, and the tritium systems. Remote maintenance requirements both inside the vacuum vessel and around the periphery of the machine were identified as the principal areas where Canadian resources could effectively provide an input, initially in requirement definition, concept evaluation and feasibility design, and subsequently in detailed design and manufacture. Support requirements were identified in such areas as the mock-up facility and a variety of planning studies relating to reliability, availability, and staff training. Specific tasks are described which provide an important data base to the facility's remote handling requirements. Canadian involvement in the areas is suggested where expertise exists and support for the remote handling work is warranted. Reliability, maintenance operations, inspection strategy and decommissioning are suggested for study. Several specific components are singled out as needing development

  3. Spent fuel and fuel pool component integrity. Annual report, FY 1979

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

  4. Classification of jet fuel properties by near-infrared spectroscopy using fuzzy rule-building expert systems and support vector machines.

    Science.gov (United States)

    Xu, Zhanfeng; Bunker, Christopher E; Harrington, Peter de B

    2010-11-01

    Monitoring the changes of jet fuel physical properties is important because fuel used in high-performance aircraft must meet rigorous specifications. Near-infrared (NIR) spectroscopy is a fast method to characterize fuels. Because of the complexity of NIR spectral data, chemometric techniques are used to extract relevant information from spectral data to accurately classify physical properties of complex fuel samples. In this work, discrimination of fuel types and classification of flash point, freezing point, boiling point (10%, v/v), boiling point (50%, v/v), and boiling point (90%, v/v) of jet fuels (JP-5, JP-8, Jet A, and Jet A1) were investigated. Each physical property was divided into three classes, low, medium, and high ranges, using two evaluations with different class boundary definitions. The class boundaries function as the threshold to alarm when the fuel properties change. Optimal partial least squares discriminant analysis (oPLS-DA), fuzzy rule-building expert system (FuRES), and support vector machines (SVM) were used to build the calibration models between the NIR spectra and classes of physical property of jet fuels. OPLS-DA, FuRES, and SVM were compared with respect to prediction accuracy. The validation of the calibration model was conducted by applying bootstrap Latin partition (BLP), which gives a measure of precision. Prediction accuracy of 97 ± 2% of the flash point, 94 ± 2% of freezing point, 99 ± 1% of the boiling point (10%, v/v), 98 ± 2% of the boiling point (50%, v/v), and 96 ± 1% of the boiling point (90%, v/v) were obtained by FuRES in one boundaries definition. Both FuRES and SVM obtained statistically better prediction accuracy over those obtained by oPLS-DA. The results indicate that combined with chemometric classifiers NIR spectroscopy could be a fast method to monitor the changes of jet fuel physical properties.

  5. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

    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

  6. Drilling Experiments of Dummy Fuel Rods Using a Mock-up Drilling Device and Detail Design of Device for Drilling of Irradiated Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Yong; Lee, H. K.; Chun, Y. B.; Park, S. J.; Kim, B. G

    2007-07-15

    KAERI are developing the safety evaluation method and the analysis technology for high burn-up nuclear fuel rod that is the project, re-irradiation for re-instrumented fuel rod. That project includes insertion of a thermocouple in the center hole of PWR nuclear fuel rod with standard burn-up, 3,500{approx}4,000MWD/tU and then inspection of the nuclear fuel rod's heat performance during re-irradiation. To re-fabricate fuel rod, two devices are needed such as a drilling machine and a welding machine. The drilling machine performs grinding a center hole, 2.5 mm in diameter and 50 mm in depth, for inserting a thermocouple. And the welding machine is used to fasten a end plug on a fuel rod. Because these two equipment handle irradiated fuel rods, they are operated in hot cell blocked radioactive rays. Before inserting any device into hot cell, many tests with that machine have to be conducted. This report shows preliminary experiments for drilling a center hole on dummy of fuel rods and optimized drilling parameters to lessen operation time and damage of diamond dills. And the design method of a drilling machine for irradiated nuclear fuel rods and detail design drawings are attached.

  7. SRV-automatic handling device

    International Nuclear Information System (INIS)

    Yamada, Koji

    1987-01-01

    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.)

  8. Sealed can of spent fuel

    International Nuclear Information System (INIS)

    Suzuki, Yasuyuki.

    1976-01-01

    Object: To provide a seal plug cover with a gripping portion fitted to a canning machine and a gripping portion fitted to a gripper of the same configuration as a fuel body for handling the fuel body so as to facilitate the handling work. Structure: A sealed can comprises a vessel and a seal plug cover, said cover being substantially in the form of a bottomed cylinder, which is slipped on the vessel and air-tightly secured by a fastening bolt between it and a flange. The spent fuel body is received into the vessel together with coolant during the step of canning operation. Said seal plug cover has two gripping portions, one for opening and closing the plug cover of the canning machine as an exclusive use member, the other being in the form of a hook-shaped peripheral groove, whereby the gripping portions may be effectively used using the same gripper when the spent fuel body is transported while being received in the sealed can or when the fuel body is removed from the sealed can. (Kawakami, Y.)

  9. Stress analysis on the valve of the rotating shield, coupled with fuel element loading-unloading machine in a PWR pressure vessel

    International Nuclear Information System (INIS)

    Albuquerque, L.B. de; Jesus Miranda, C.A. de.

    1992-01-01

    A finite element static analysis was performed with the valve of the Rotating Shield (RS) which is coupled with the Fuel. Element Loading-Unloading Machine under OBE earthquake. The applied leads were obtained from a previous seismic analysis with the response spectrum method of the MTC under OBE load. A 3-D model with shell elements was developed for the valve body and for a part of the RS. The ANSYS program, version 4.4 A, was used. The two main scopes of this work were to verify the valve stresses and the functionality of its moving parts during the earthquake. (author)

  10. Nuclear fuel preheating system

    International Nuclear Information System (INIS)

    Andrea, C.

    1975-01-01

    A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

  11. Safe handling of kilogram amounts of fuel-grade plutonium and of gram amounts of plutonium-238, americium-241 and curium-244

    International Nuclear Information System (INIS)

    Louwrier, K.P.; Richter, K.

    1976-01-01

    During the past 10 years about 600 glove-boxes have been installed at the Institute for Transuranium Elements at Karlsruhe. About 80% of these glove-boxes have been designed and equipped for handling 100-g to 1-kg amounts of 239 Pu containing 8-12% 240 Pu (low-exposure plutonium). A small proportion of the glove-boxes is equipped with additional shielding in the form of lead sheet or lead glass for work with recycled plutonium. In these glove-boxes gram-amounts of 241 Am have also been handled for preparation of Al-Am targets using tongs and additional shielding inside the glove-boxes themselves. Water- and lead-shielded glove-boxes equipped with telemanipulators have been installed for routine work with gram-amounts of 241 Am, 243 Am and 244 Cm. A prediction of the expected radiation dose for the personnel is difficult and only valid for a preparation procedure with well-defined preparation steps, owing to the fact that gamma dose-rates depend strongly upon proximity and source seize. Gamma radiation dose measurements during non-routine work for 241 Am target preparation showed that handling of gram amounts leads to a rather high irradiation dose for the personnel, despite lead or steel glove-box shielding and shielding within the glove-boxes. A direct glove-hand to americium contact must be avoided. For all glove-handling of materials with gamma radiation an irradiation control of the forearms of the personnel by, for example, thermoluminescence dosimeters is necessary. Routine handling of americium and curium should be executed with master-slave equipment behind neutron and gamma shielding. (author)

  12. Contact-type displacement measuring mechanism for fuel assembly in reactor

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Ko, Kuniaki.

    1995-01-01

    The measuring mechanism of the present invention, which is used in a lmfbr type reactor, is suspended by a gripper of a fuel handing machine, and it comprises a combination of a displacement amount measuring jig allowed to be inserted into a handling head of a fuel assembly and a displacement amount measuring ring disposed at the lower portion in the handling head. The displacement amount measuring jig has a structure comprising a releasable handle and a columnar or cylindrical measuring portion allowable to be inserted into the handling head formed at the lower portion of the handle, which are connected with each other. When an interference (contact) occurred between the displacement amount measuring jig and the stepwise displacement amount measuring ring during the measurement, change of load and a phenomenon that the fuel handing machine can not be lowered are recognized, so that core displacement amount can be recognized based on the stroke of the gripper portion. Then, remote measurement is possible for displacement and deformation of the fuel assembly in the reactor container, and the measurement can be conducted by the same procedures and in the same period of time as in a case of ordinary fuel exchange operation. A flow channel for coolants passing through the fuel assembly can be ensured, thereby enabling to measure the amount of core displacement which is closer to an actual value in the reactor. (N.H.)

  13. Fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A new fuel can with a loose bottom and head is described. The fuel bar is attached to the loose bottom and head with two grid poles keeping the distance between bottom and head. A bow-shaped handle is attached to the head so that the fuel bar can be lifted from the can

  14. Implications of manufacturing deviations on fuel performance

    International Nuclear Information System (INIS)

    Chellapandi, P.; Clement Ravi Chandar, S.; Chetal, S.C.; Baldev Raj

    2009-01-01

    Prototype Fast Breeder Reactor (PFBR) core consists of 181 Fuel subassembly (FSA), 120 blanket SA and shielding SA of steel and B 4 C besides 9 Control Safety Rod SA and 3 Diverse Safety Rod SA. All the subassemblies (SA) stand vertically on the grid plate. PFBR FSA consists of 217 fuel pins of 2540 mm length arranged in a triangular pitch standing vertically on the rails inside a hexagonal duct. The bottom of the hexagon is screwed and welded to a cylindrical foot of length 600 mm. The SA foot has radial slots that provide an entry to the coolant sodium and also houses flow control devices. While the foot of the A handling head is welded to the top of hexagon, provides an aid for fuel handling machine to insert and withdraw the FSA from the grid plate. The length of the FSA is 4500 mm.

  15. Adaptive Control Using Fully Online Sequential-Extreme Learning Machine and a Case Study on Engine Air-Fuel Ratio Regulation

    Directory of Open Access Journals (Sweden)

    Pak Kin Wong

    2014-01-01

    Full Text Available Most adaptive neural control schemes are based on stochastic gradient-descent backpropagation (SGBP, which suffers from local minima problem. Although the recently proposed regularized online sequential-extreme learning machine (ReOS-ELM can overcome this issue, it requires a batch of representative initial training data to construct a base model before online learning. The initial data is usually difficult to collect in adaptive control applications. Therefore, this paper proposes an improved version of ReOS-ELM, entitled fully online sequential-extreme learning machine (FOS-ELM. While retaining the advantages of ReOS-ELM, FOS-ELM discards the initial training phase, and hence becomes suitable for adaptive control applications. To demonstrate its effectiveness, FOS-ELM was applied to the adaptive control of engine air-fuel ratio based on a simulated engine model. Besides, controller parameters were also analyzed, in which it is found that large hidden node number with small regularization parameter leads to the best performance. A comparison among FOS-ELM and SGBP was also conducted. The result indicates that FOS-ELM achieves better tracking and convergence performance than SGBP, since FOS-ELM tends to learn the unknown engine model globally whereas SGBP tends to “forget” what it has learnt. This implies that FOS-ELM is more preferable for adaptive control applications.

  16. Innovative automation solutions applied to nuclear fuel production and inspection processes

    International Nuclear Information System (INIS)

    Vas, Ananth

    2012-01-01

    The nuclear industry in India is slated for fast paced growth in the coming years, with a great focus on increasing the capacity for producing, inspecting and finally reprocessing of nuclear fuel. Modern techniques of industrial automation such as robotics, machine vision and laser based systems have been deployed extensively to improve the productivity and output of existing and future installations, particularly for the fuel handling stages mentioned

  17. DNA-based machines.

    Science.gov (United States)

    Wang, Fuan; Willner, Bilha; Willner, Itamar

    2014-01-01

    The base sequence in nucleic acids encodes substantial structural and functional information into the biopolymer. This encoded information provides the basis for the tailoring and assembly of DNA machines. A DNA machine is defined as a molecular device that exhibits the following fundamental features. (1) It performs a fuel-driven mechanical process that mimics macroscopic machines. (2) The mechanical process requires an energy input, "fuel." (3) The mechanical operation is accompanied by an energy consumption process that leads to "waste products." (4) The cyclic operation of the DNA devices, involves the use of "fuel" and "anti-fuel" ingredients. A variety of DNA-based machines are described, including the construction of "tweezers," "walkers," "robots," "cranes," "transporters," "springs," "gears," and interlocked cyclic DNA structures acting as reconfigurable catenanes, rotaxanes, and rotors. Different "fuels", such as nucleic acid strands, pH (H⁺/OH⁻), metal ions, and light, are used to trigger the mechanical functions of the DNA devices. The operation of the devices in solution and on surfaces is described, and a variety of optical, electrical, and photoelectrochemical methods to follow the operations of the DNA machines are presented. We further address the possible applications of DNA machines and the future perspectives of molecular DNA devices. These include the application of DNA machines as functional structures for the construction of logic gates and computing, for the programmed organization of metallic nanoparticle structures and the control of plasmonic properties, and for controlling chemical transformations by DNA machines. We further discuss the future applications of DNA machines for intracellular sensing, controlling intracellular metabolic pathways, and the use of the functional nanostructures for drug delivery and medical applications.

  18. Remote handling in reprocessing plants

    International Nuclear Information System (INIS)

    Streiff, G.

    1984-01-01

    Remote control will be the rule for maintenance in hot cells of future spent fuel reprocessing plants because of the radioactivity level. New handling equipments will be developed and intervention principles defined. Existing materials, recommendations for use and new manipulators are found in the PMDS' documentation. It is also a help in the choice and use of intervention means and a guide for the user [fr

  19. Feasibility demonstration of using wire electrical-discharge machining, abrasive flow honing, and laser spot welding to manufacture high-precision triangular-pitch Zircaloy-4 fuel-rod-support grids

    International Nuclear Information System (INIS)

    Horwood, W.A.

    1982-05-01

    Results are reported supporting the feasibility of manufacturing high precision machined triangular pitch Zircaloy-4 fuel rod support grids for application in water cooled nuclear power reactors. The manufacturing processes investigated included wire electrical discharge machining of the fuel rod and guide tube cells in Zircaloy plate stock to provide the grid body, multistep pickling of the machined grid to provide smooth and corrosion resistant surfaces, and laser welding of thin Zircaloy cover plates to both sides of the grid body to capture separate AM-350 stainless steel insert springs in the grid body. Results indicated that dimensional accuracy better than +- 0.001 and +- 0.002 inch could be obtained on cell shape and position respectively after wire EDM and surface pickling. Results on strength, corrosion resistance, and internal quality of laser spot welds are provided

  20. On the Generalized Correlation Equation of Welding Current for the Tig Welding Machine Used in Nuclear Fuel Fabrication

    International Nuclear Information System (INIS)

    Umar, Efrizon

    1995-01-01

    In nuclear fuel fabrication, welding plays a very important role to join the end cap to the tube. In order to determine the welding current in TIG welding process for various materials, weld geometries and welding rates, the correlation between the welding current and the other parameters are needed. This paper presents the correlation of those parameters mentioned above. The proposed correlation was tested and produced satisfactory results. (author). 8 refs., 2 tabs., 2 figs

  1. Nuclear reactor machine refuelling system

    International Nuclear Information System (INIS)

    Cashen, W.S.; Erwin, D.

    1977-01-01

    Part of an on-line fuelling machine for a CANDU pressure-tube reactor is described. The present invention provides a refuelling machine wherein the fuelling components, including the fuel carrier and the closure adapter, are positively positioned and retained within the machine magazine or positively secured to the machine charge tube head, and cannot be accidentally disengaged as in former practice. The positive positioning devices include an arcuate keeper plate. Simplified hooked fingers are used. (NDH)

  2. Human-machine interfaces for teleoperators: an overview of research and development at the Oak Ridge National Laboratory. Consolidated Fuel Reprocessing Program

    International Nuclear Information System (INIS)

    Draper, J.V.; Feldman, M.J.

    1985-01-01

    This paper surveys the contributions of human factors to the mission of the Remote Control Engineering (RCE) task over the last six years. These contributions can be divided into two areas, research efforts and design efforts. Some of the topics covered in human factors research are manipulator comparisons, investigation of viewing system characteristics, research into the effects of force reflection, and studies of crew size and task allocation. In the area of component design the human factors group was primarily responsible for the conceptual design of the Advanced Integrated Maintenance System (AIMS) control room, including all operator work stations and overall control room architecture. The human factors group also contributed to the design of the AIMS master controller handle. Recent research at the RCE task has centered on comparison of manipulator systems. This research was planned and conducted by the human factors group and other ORNL personnel. The research is aimed at evaluating three important characteristics of manipulator systems: system dynamics, force feedback, and human-machine interface

  3. Research on plant of metal fuel fabrication using casting process (2)

    International Nuclear Information System (INIS)

    Senda, Yasuhide; Yamada, Seiya

    2005-02-01

    In this research work for the metal fuel fabrication system (38 tHM/y), the studies of the concept of the main process equipments were performed based on the previous studies on the process design and the quality control system design. In this study the handling equipment of the products were also designed, according to these designs the handling periods were evaluated. Consequently the numbers of the equipments were assessed taking into account for the method of the blending the fuel composition. (1) Structural concept design of the major equipments, the fuel handling machine and the gravimetries in the main fabrication process. The structural concept were designed for the fuel composition blending equipment, the fuel pin assembling equipment, the sodium bonding equipment, the handling equipment for fuel slug palettes, the handling equipment for fuel pins and the gravimetries. (2) Re-assessment of the numbers of the equipments taking account of the handling periods. Based on the results of item (1) the periods were evaluated for the fuel slug and pin handling. Processing time of demolder is short, then the number of it is increased to two. Three vehicles are also added to transfer the slugs and a heel smoothly. (3) Design of the buffer storages. The buffer storages among the equipments were designed through the comparison of the process speed between the equipments taking into account for the handling periods. The required amount of the structural parts (for example cladding materials) was assessed for the buffer in the same manner and the amount of the buffer facilities were optimized. (author)

  4. Robotics and remote handling in the nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This book presents the papers given at a conference on the use of remote handling equipment in nuclear facilities. Topics considered at the conference included dose reduction, artificial intelligence in nuclear plant maintenance, robotic welding, uncertainty covariances, reactor operation and inspection, reactor maintenance and repair, uranium mining, fuel fabrication, reactor component manufacture, irradiated fuel and radioactive waste management, and radioisotope handling.

  5. Radiological safety aspects of handling plutonium

    International Nuclear Information System (INIS)

    Sundararajan, A.R.

    2016-01-01

    Department of Atomic Energy in its scheme of harnessing the nuclear energy for electrical power generation and strategic applications has given a huge role to utilization of plutonium. In the power production programme, fast reactors with plutonium as fuel are expected to play a major role. This would require establishing fuel reprocessing plants to handle both thermal and fast reactor fuels. So in the nuclear fuel cycle facilities variety of chemical, metallurgical, mechanical operations have to be carried out involving significant inventories of "2"3"9 Pu and associated radionuclides. Plutonium is the most radiotoxic radionuclide and therefore any facility handling it has to be designed and operated with utmost care. Two problems of major concern in the protection of persons working in plutonium handling facilities are the internal exposure to the operating personnel from uptake of plutonium and transplutonic nuclides as they are highly radiotoxic and the radiation exposure of hands and eye lens during fuel fabrication operations especially while handling recycled high burn up plutonium. In view of the fact that annual limit for intake is very small for "2"3"9Pu and its radiation emission characteristics are such that it is a huge challenge for the health physicists to detect Pu in air and in workers. This paper discusses the principles and practices followed in providing radiological surveillance to workers in plutonium handling areas. The challenges in protecting the workers from receiving exposures to hands and eye lens in handling high burn up plutonium are also discussed. The sites having Pu fuel cycle facilities should have trained medical staff to handle cases involving excessive intake of plutonium. (author)

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1970-01-01

    Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

  7. Use of micro gas chromatography in the fuel cycle of fusion reactors

    International Nuclear Information System (INIS)

    Laesser, R.; Gruenhagen, S.; Kawamura, Y.

    2003-01-01

    Various analytical techniques exist to determine the compositions of gases handled in the fuel cycle of future fusion machines. Gas chromatography was found to be the most appropriate method. The main disadvantages of conventional gas chromatography were the long retention times for the heavy hydrogen species of >30 min. Recent progress in the development of micro-gas chromatography has reduced these retention times to ∼3 min. The usefulness of micro-gas chromatography for the analysis of hydrogen and impurity gas mixtures in the fuel cycle of future fusion machines is presented and the advantages and drawbacks are discussed

  8. Chipping machines: disc and drum energy requirements

    Directory of Open Access Journals (Sweden)

    Alessio Facello

    2013-09-01

    Full Text Available Air pollution and fossil fuel reserves exhaustion are increasing the importance of the biomass-derived products, in particular wood, as source of clean and renewable energy for the production of electricity or steam. In order to improve the global efficiency and the entire production chain, we have to evaluate the energetic aspects linked to the process of transformation, handling and transport of these materials. This paper reports results on a comparison between two chippers of similar size using different cutting technology: disc and drum tool respectively. During trials, fuel consumption, PTO torque and speed, processing time and weight of processed material were recorded. Power demand, fuel consumption, specific energy and productivity were computed. The machine was fed with four different feedstock types (chestnut logs, poplar logs, poplar branches, poplar sawmill residues. 15 repetitions for each combination of feedstock-tool were carried out. The results of this study show that the disc tool requires, depending on the processed material, from 12 to 18% less fuel per unit of material processed than the drum tool, and consequently, from 12 to 16% less specific energy. In particular, the highest difference between tools was found in branches processing whereas the smallest was in poplar logs. Furthermore the results of the investigation indicate, that, in testing conditions, the productivity of drum tool is higher (8% than disc tool.

  9. The effective and dust free receiving station and handling for the low calorific value solid fuels; Tehokas ja poelytoen seospolttoaineiden vastaanottoasema sekae kaesittely- ja kuljetinjaerjestelmae

    Energy Technology Data Exchange (ETDEWEB)

    Nojonen, O [Finntech Oy, Espoo (Finland); Jaervinen, T [VTT Energy, Jyvaeskylae (Finland). Energy Use

    1996-12-31

    The aim of the project was to get higher automatization level and improve dust preventing in solid fuel receiving stations. There are two general types of receiving stations in the Finnish power plants: large unloading stations for the side tipping devices of trucks and small ones for the rear tipping devices of trucks. In the first ones the trucks empty their load (approx. 100 m{sup 3} loose bulk density) divided by hauling unit and trailer within few (5-10) minutes into a rectangular box, which depth is approx. 3 m. The discharging causes a strong counter current air and dust flow (20-40 m{sup 3}/s) upwards and the dust will easily spread out all over the station. In the second ones the discharging takes place from the rear of truck and trailer using loadspace conveyor of the vehicle within 20 minutes. The material falls on a (belt) conveyor, which is on the floor level. The problems in side tipping system are connected with dust and surplus time, which is needed for the preparation and completion of unloading and sampling. For the fine dust control has also been tested water spray (fog) nozzles and tried to utilise a settling chamber for the dust stream. Also the using the settling chamber equipped with air suction connections and cyclone separation may be effective. In the rear tipping system are also applied bag filters because of smaller air quantities. The rapid truck positioning and control in the receiving station is one of the presupposition for the fast and accurate unloading. This can be done using ultrasonic sensing methods. Also the ensuring of accurate discharging can be based on the modern technology. One of the basic things is the enlarging of video control (CCD -cameras)

  10. The effective and dust free receiving station and handling for the low calorific value solid fuels; Tehokas ja poelytoen seospolttoaineiden vastaanottoasema sekae kaesittely- ja kuljetinjaerjestelmae

    Energy Technology Data Exchange (ETDEWEB)

    Nojonen, O. [Finntech Oy, Espoo (Finland); Jaervinen, T. [VTT Energy, Jyvaeskylae (Finland). Energy Use

    1995-12-31

    The aim of the project was to get higher automatization level and improve dust preventing in solid fuel receiving stations. There are two general types of receiving stations in the Finnish power plants: large unloading stations for the side tipping devices of trucks and small ones for the rear tipping devices of trucks. In the first ones the trucks empty their load (approx. 100 m{sup 3} loose bulk density) divided by hauling unit and trailer within few (5-10) minutes into a rectangular box, which depth is approx. 3 m. The discharging causes a strong counter current air and dust flow (20-40 m{sup 3}/s) upwards and the dust will easily spread out all over the station. In the second ones the discharging takes place from the rear of truck and trailer using loadspace conveyor of the vehicle within 20 minutes. The material falls on a (belt) conveyor, which is on the floor level. The problems in side tipping system are connected with dust and surplus time, which is needed for the preparation and completion of unloading and sampling. For the fine dust control has also been tested water spray (fog) nozzles and tried to utilise a settling chamber for the dust stream. Also the using the settling chamber equipped with air suction connections and cyclone separation may be effective. In the rear tipping system are also applied bag filters because of smaller air quantities. The rapid truck positioning and control in the receiving station is one of the presupposition for the fast and accurate unloading. This can be done using ultrasonic sensing methods. Also the ensuring of accurate discharging can be based on the modern technology. One of the basic things is the enlarging of video control (CCD -cameras)

  11. The health physics of installations for decladding irradiated fuels or for handling radio-elements at Marcoule; La radioprotection des installations de degainage des combustibles irradies et des radio-elements a Marcoule

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J; Guillermin, P; Mallet, P [Commissariat a l' Energie Atomique, Centre de Production de Plutonium de Marcoule, 30 - Chusclan (France)

    1966-07-01

    Radiation protection practices for handling reactor fuel elements are described. Elements of considerable specific radioactivity are handled under water. A study was made of water filtration and of ventilation in the building. The installations are divided up into zones depending on the radioactive risks, and the radiation level atmospheric contamination are the object of a systematic control at various points. A description is given of all aspects of health physics which have been encountered during six years: storage, transfer of radioactive material; decladding, rolling, waste disposal, specialized operations, installations in operation and at rest, and transport. In spite of the gradual increase in the activity of these installations, the total doses received by the personnel have hardly altered and the number of cases of physical contamination has diminished. (authors) [French] Dans ces installations, se manipulent sous l'eau des elements a radioactivite specifique considerable. La filtration de l'eau, la ventilation ont ete particulierement etudiees. L'ensemble a ete divise en lieux classes en fonction des risques radioactifs et des appareils controlent en permanence l'irradiation et la contamination atmospherique en certains points. Tous les aspects de la radioprotection resultant de six annees d'experience relatifs: au stockage, au deconteneurage, au degainage, au laminage, a l'evacuation des residus, aux travaux particuliers, installations en marche et a l'arret, et aux transports sont successivement decrits. Malgre l'accroissement progressif de l'activite de cet ensemble, les doses integrees par le personnel n'ont pratiquement pas augmente et le nombre des cas de contamination corporelle a diminue. (auteurs)

  12. Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

    International Nuclear Information System (INIS)

    Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

    2010-01-01

    A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

  13. Machine Shop Grinding Machines.

    Science.gov (United States)

    Dunn, James

    This curriculum manual is one in a series of machine shop curriculum manuals intended for use in full-time secondary and postsecondary classes, as well as part-time adult classes. The curriculum can also be adapted to open-entry, open-exit programs. Its purpose is to equip students with basic knowledge and skills that will enable them to enter the…

  14. Fuel assembly guide tube

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations

  15. Spent fuel canister docking station

    International Nuclear Information System (INIS)

    Suikki, M.

    2006-01-01

    The working report for the spent fuel canister docking station presents a design for the operation and structure of the docking equipment located in the fuel handling cell for the spent fuel in the encapsulation plant. The report contains a description of the basic requirements for the docking station equipment and their implementation, the operation of the equipment, maintenance and a cost estimate. In the designing of the equipment all the problems related with the operation have been solved at the level of principle, nevertheless, detailed designing and the selection of final components have not yet been carried out. In case of defects and failures, solutions have been considered for postulated problems, and furthermore, the entire equipment was gone through by the means of systematic risk analysis (PFMEA). During the docking station designing we came across with needs to influence the structure of the actual disposal canister for spent nuclear fuel, too. Proposed changes for the structure of the steel lid fastening screw were included in the report. The report also contains a description of installation with the fuel handling cell structures. The purpose of the docking station for the fuel handling cell is to position and to seal the disposal canister for spent nuclear fuel into a penetration located on the cell floor and to provide suitable means for executing the loading of the disposal canister and the changing of atmosphere. The designed docking station consists of a docking ring, a covering hatch, a protective cone and an atmosphere-changing cap as well as the vacuum technology pertaining to the changing of atmosphere and the inert gas system. As far as the solutions are concerned, we have arrived at rather simple structures and most of the actuators of the system are situated outside of the actual fuel handling cell. When necessary, the equipment can also be used for the dismantling of a faulty disposal canister, cut from its upper end by machining. The

  16. Analysis and simulation of straw fuel logistics

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Daniel [Swedish Univ. of Agricultural Sciences, Uppsala (Sweden). Dept. of Agricultural Engineering

    1998-12-31

    Straw is a renewable biomass that has a considerable potential to be used as fuel in rural districts. This bulky fuel is, however, produced over large areas and must be collected during a limited amount of days and taken to the storages before being ultimately transported to heating plants. Thus, a well thought-out and cost-effective harvesting and handling system is necessary to provide a satisfactory fuel at competitive costs. Moreover, high-quality non-renewable fuels are used in these operations. To be sustainable, the energy content of these fuels should not exceed the energy extracted from the straw. The objective of this study is to analyze straw as fuel in district heating plants with respect to environmental and energy aspects, and to improve the performance and reduce the costs of straw handling. Energy, exergy and emergy analyses were used to assess straw as fuel from an energy point of view. The energy analysis showed that the energy balance is 12:1 when direct and indirect energy requirements are considered. The exergy analysis demonstrated that the conversion step is ineffective, whereas the emergy analysis indicated that large amounts of energy have been used in the past to form the straw fuel (the net emergy yield ratio is 1.1). A dynamic simulation model, called SHAM (Straw HAndling Model), has also been developed to investigate handling of straw from the fields to the plant. The primary aim is to analyze the performance of various machinery chains and management strategies in order to reduce the handling costs and energy needs. The model, which is based on discrete event simulation, takes both weather and geographical conditions into account. The model has been applied to three regions in Sweden (Svaloev, Vara and Enkoeping) in order to investigate the prerequisites for straw harvest at these locations. The simulations showed that straw has the best chances to become a competitive fuel in south Sweden. It was also demonstrated that costs can be

  17. Repository waste-handling operations, 1998

    International Nuclear Information System (INIS)

    Cottam, A.E.; Connell, L.

    1986-04-01

    The Civilian Radioactive Waste Management Program Mission Plan and the Generic Requirements for a Mined Geologic Disposal System state that beginning in 1998, commercial spent fuel not exceeding 70,000 metric tons of heavy metal, or a quantity of solidified high-level radioactive waste resulting from the reprocessing of such a quantity of spent fuel, will be shipped to a deep geologic repository for permanent storage. The development of a waste-handling system that can process 3000 metric tons of heavy metal annually will require the adoption of a fully automated approach. The safety and minimum exposure of personnel will be the prime goals of the repository waste handling system. A man-out-of-the-loop approach will be used in all operations including the receipt of spent fuel in shipping casks, the inspection and unloading of the spent fuel into automated hot-cell facilities, the disassembly of spent fuel assemblies, the consolidation of fuel rods, and the packaging of fuel rods into heavy-walled site-specific containers. These containers are designed to contain the radionuclides for up to 1000 years. The ability of a repository to handle more than 6000 pressurized water reactor spent-fuel rods per day on a production basis for approximately a 23-year period will require that a systems approach be adopted that combines space-age technology, robotics, and sophisticated automated computerized equipment. New advanced inspection techniques, maintenance by robots, and safety will be key factors in the design, construction, and licensing of a repository waste-handling facility for 1998

  18. Natural Nano-Machines

    Indian Academy of Sciences (India)

    Administrator

    transport, ion pump, ATP syn- thase. A popularized ..... gas. A lice: I could not understand how A T P m olecules serve as fuels for m olecular m achines. ..... [16] V Balzani, M Venturi and A Credi, Molecular Devices and Machines: a Journey into ...

  19. Low-cost concepts for dry transfer of spent fuel and waste between storage and transportation casks

    International Nuclear Information System (INIS)

    Schneider, K.L.

    1984-01-01

    The federal government may provide interim storage for spent fuel from commercial nuclear power reactors that have used up their available storage capacity. One of the leading candidate concepts for this interim storage is to place spent fuel in large metal shielding casks. The Federal Interim Storage (FIS) site may not have the capability to transfer spent fuel from transportation casks to storage casks and vice versa. Thus, there may be an incentive to construct a relatively inexpensive but reliable intercask transfer system for use at an FIS site. This report documents the results of a preliminary study of preconceptual design and analysis of four intercask transfer concepts. The four concepts are: a large shielded cylindrical turntable that contains an integral fuel handling machine (turntable concept); a shielded fuel handling machine under which shipping and storage casks are moved horizontally (shuttle concept); a small hot cell containing equipment for transferring fuel betwee shipping and storage casks (that enter and leave the cell on carts) in a bifurcated trench (trench concept) and a large hot cell, shielded by an earthen berm, that houses equipment for handling fuel between casks that enter and leave the cell on a single cart (igloo concept). Information derived for each of the concepts is operating, capital and relocation costs; implementation and relocation time requirements; and overall characteristics

  20. Handling and disposing of radioactive waste

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1983-01-01

    Radioactive waste has been separated by definition into six categories. These are: commercial spent fuel; high-level wastes; transuranium waste; low-level wastes; decommissioning and decontamination wastes; and mill tailings and mine wastes. Handling and disposing of these various types of radioactive wastes are discussed briefly

  1. Biodiesel Handling and Use Guide (Fifth Edition)

    Energy Technology Data Exchange (ETDEWEB)

    Alleman, T.L.; McCormick, R.L.; Christensen, E.D.; Fioroni, G.; Moriarty. K.; Yanowitz, J.

    2016-11-08

    This document is a guide for those who blend, distribute, and use biodiesel and biodiesel blends. It provides basic information on the proper and safe use of biodiesel and biodiesel blends in engines and boilers, and is intended to help fleets, individual users, blenders, distributors, and those involved in related activities understand procedures for handling and using biodiesel fuels.

  2. Harvesting and handling agricultural residues for energy

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, B.M.; Summer, H.R.

    1986-05-01

    Significant progress in understanding the needs for design of agricultural residue collection and handling systems has been made but additional research is required. Recommendations are made for research to (a) integrate residue collection and handling systems into general agricultural practices through the development of multi-use equipment and total harvest systems; (b) improve methods for routine evaluation of agricultural residue resources, possibly through remote sensing and image processing; (c) analyze biomass properties to obtain detailed data relevant to engineering design and analysis; (d) evaluate long-term environmental, social, and agronomic impacts of residue collection; (e) develop improved equipment with higher capacities to reduce residue collection and handling costs, with emphasis on optimal design of complete systems including collection, transportation, processing, storage, and utilization; and (f) produce standard forms of biomass fuels or products to enhance material handling and expand biomass markets through improved reliability and automatic control of biomass conversion and other utilization systems. 118 references.

  3. Treatment of Markup in Statistical Machine Translation

    OpenAIRE

    Müller, Mathias

    2017-01-01

    We present work on handling XML markup in Statistical Machine Translation (SMT). The methods we propose can be used to effectively preserve markup (for instance inline formatting or structure) and to place markup correctly in a machine-translated segment. We evaluate our approaches with parallel data that naturally contains markup or where markup was inserted to create synthetic examples. In our experiments, hybrid reinsertion has proven the most accurate method to handle markup, while alignm...

  4. Remote Inspection, Measurement and Handling for LHC

    CERN Document Server

    Kershaw, K; Coin, A; Delsaux, F; Feniet, T; Grenard, J L; Valbuena, R

    2007-01-01

    Personnel access to the LHC tunnel will be restricted to varying extents during the life of the machine due to radiation, cryogenic and pressure hazards. The ability to carry out visual inspection, measurement and handling activities remotely during periods when the LHC tunnel is potentially hazardous offers advantages in terms of safety, accelerator down time, and costs. The first applications identified were remote measurement of radiation levels at the start of shut-down, remote geometrical survey measurements in the collimation regions, and remote visual inspection during pressure testing and initial machine cool-down. In addition, for remote handling operations, it will be necessary to be able to transmit several real-time video images from the tunnel to the control room. The paper describes the design, development and use of a remotely controlled vehicle to demonstrate the feasibility of meeting the above requirements in the LHC tunnel. Design choices are explained along with operating experience to-dat...

  5. How to Handle Abuse

    Science.gov (United States)

    ... Handle Abuse KidsHealth / For Kids / How to Handle Abuse What's in this article? Tell Right Away How Do You Know Something Is Abuse? ... babysitter, teacher, coach, or a bigger kid. Child abuse can happen anywhere — at ... building. Tell Right Away A kid who is being seriously hurt ...

  6. Grain Handling and Storage.

    Science.gov (United States)

    Harris, Troy G.; Minor, John

    This text for a secondary- or postecondary-level course in grain handling and storage contains ten chapters. Chapter titles are (1) Introduction to Grain Handling and Storage, (2) Elevator Safety, (3) Grain Grading and Seed Identification, (4) Moisture Control, (5) Insect and Rodent Control, (6) Grain Inventory Control, (7) Elevator Maintenance,…

  7. Door valve for fuel handling path

    International Nuclear Information System (INIS)

    Makishima, Katsuhiko.

    1969-01-01

    A door valve is provided which seals cover gas from a liquid metal cooled reactor without leakage therefrom. A threaded shaft is screwed into a heavy box press which is packed with lead. The shaft is adapted to be rotated by an electric motor or a manually operated wheel which is disposed outside of the door valve. A valve plate is suspended from the box press by four guide wheels mounted thereon. The guide wheels are fitted into inclined guide grooves formed at the valve plate and into grooved formed in the inner wall of a valve casing. A locking ball is provided at each side of the valve plate. In operation the shaft rotates and travels to permit the box press and the valve plate to move into the door valve casing, thus releasing the locking balls. The valve plate does not contact the bottom of the casing. When the box press reaches the home position, the valve plate is carried on the valve opening, and the box press presses the valve plate to increase the tightness. The valve plate does not suffer wear as it does not slide over other parts. (Yamaguchi, T.)

  8. Door valve for fuel handling path

    International Nuclear Information System (INIS)

    Makishima, Katsuhiko.

    1969-01-01

    A door valve is provided which seals cover gas from a liquid metal cooled reactor without leakage therefrom. A threaded shaft is screwed into a heavy box press which is packed with lead. The shaft is adapted to be rotated by an electric motor or a manually operated wheel which is disposed outside of the door valve. From the box press a valve plate is suspended by four linkage bars, one for each corner. Each linkage bar is provided with two wheels which are respectively mounted at the connections with the box press and the valve plate. The wheels are carried on the horizontal grooves formed in a door valve casing. In operation the shaft rotates and travels to permit the box press and the valve plate to move into the door valve casing while the valve plate does not contact the casing. When the box press reaches the home position, the wheels drop into the recesses which are disposed at the ends of the grooves, the valve plate is carried on the valve opening, and the box press presses the valve plate to increase the tightness. The valve plate does not suffer wear as it does not over other parts. (Yamaguchi, T.)

  9. Study of technical and financial pre-feasibility for the handling and sale, as fuel in cement industries, of petroleum coke produced at a petroleum refinery located in Moin, Limon

    International Nuclear Information System (INIS)

    Corrales Suarez, Jeffry

    2013-01-01

    A study of the technical and financial pre-feasibility is realized for the handling, conditioning and sale as fuel in cement industries of petroleum coke, that is produced in the project of extension and modernization of the refinery of RECOPE in Moin, Limon. Daily production has been of 570 metric tons. The market study has determined a demand of approximately 120 000 metric tons in the country and in the region. The total production of coke has been possible place it at national and regional level, due to lack of another producer. The coke is recommended to crush it to decrease the size of the particles for the conditioning, handling and to facilitate their manipulation. Conveyor belts are used for handling on the inside of the refinery. The coke is transported in trucks type tanker on the exterior of the refinery. The moisture content is reduced by dryer vibratory of fluidized bed to 5% of moisture to guarantee a product better quality. The product is stored under roof in ventilated facilities and with appropriate systems of security to minimize the risk of an accident. The fixed capital investment to develop the project has been of USD 3,1 millions and working capital of USD 14,4 millions. The financial evaluation is realized considering two financing models of 70% and without financing. The net present value (NPV) has been of USD -13,0 millions and the desirability index of -6,6 for the model with financing. The model without financing has obtained a NPV of USD -1,9 millions and a desirability index of -1,2. The financial profitability of the project has been very sensitive to the sales price of the coke, to the coke volume of production, to the growth percentage of the sale price and the cost of the raw material for the two models considered. The financial risk analysis is realized by the Monte Carlo method, whereupon is obtained a probability of approximately 34% in that the internal rate of return (IRR) has been higher than the cut rate and of 36% in that

  10. Safety Analysis of 'Older/Aged' Handling and Transportation Equipment for Heavy Loads, Radioactive Waste and Materials in Accordance with German Nuclear Standards KTA 3902, 3903 and 3905

    International Nuclear Information System (INIS)

    Macias, P.; Prucker, E.; Stang, W.

    2006-01-01

    The purpose of this paper is to present a general safety analysis of important handling and transportation processes and their related equipment ('load chains' consisting of cranes, load-bearing equipment and load-attaching points). This project was arranged by the responsible Bavarian ministry for environment, health and consumer protection (StMUGV) in agreement with the power plant operators of all Bavarian nuclear power plants to work out potential safety improvements. The range of the equipment (e.g. reactor building, crane, refuelling machine, load-bearing equipment and load-attaching points) covers the handling and transportation of fuel elements (e. g. with fuel flasks), heavy loads (e.g. reactor pressure vessel closure head, shielding slabs) and radioactive materials and waste (e.g. waste flasks, control elements, fuel channels, structure elements). The handling equipment was subjected to a general safety analysis taking into account the ageing of the equipment and the progress of standards. Compliance with the current valid requirements of the state of science and technology as required by German Atomic Act and particularly of the nuclear safety KTA-standards (3902, 3903 and 3905) was examined. The higher protection aims 'safe handling and transportation of heavy loads and safe handling of radioactive materials and waste' of the whole analysis are to avoid a criticality accident, the release of radioactivity and inadmissible effects on important technical equipment and buildings. The scope of the analysis was to check whether these protection aims were fulfilled for all important technical handling and transportation processes. In particularly the design and manufacturing of the components and the regulations of the handling itself were examined. (authors)

  11. Charging machine for a fast production reactor

    International Nuclear Information System (INIS)

    Artem'ev, L.N.; Kurilkin, V.V.

    1971-01-01

    Charging machine for a fast production reactor is described. The machine contains charging mechanism, mechanism for positioning fresh fuel and spent fuel assemtlies, storage drums with sockets for control rod assemtlies and collet tongs for control rods. Recharging is conducted by means of ramp channel

  12. Remote handling prospects. Computer aided remote handling

    International Nuclear Information System (INIS)

    Vertut, J.

    1984-01-01

    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 [fr

  13. Simultaneous scheduling of machines and mobile robots

    DEFF Research Database (Denmark)

    Dang, Vinh Quang; Nielsen, Izabela Ewa

    2013-01-01

    This paper deals with the problem of simultaneously scheduling machines and a number of autonomous mobile robots in a flexible manufacturing system (FMS). Besides capability of transporting materials between machines, the considered mobile robots are different from other material handling devices...

  14. Visible Machine Learning for Biomedicine.

    Science.gov (United States)

    Yu, Michael K; Ma, Jianzhu; Fisher, Jasmin; Kreisberg, Jason F; Raphael, Benjamin J; Ideker, Trey

    2018-06-14

    A major ambition of artificial intelligence lies in translating patient data to successful therapies. Machine learning models face particular challenges in biomedicine, however, including handling of extreme data heterogeneity and lack of mechanistic insight into predictions. Here, we argue for "visible" approaches that guide model structure with experimental biology. Copyright © 2018. Published by Elsevier Inc.

  15. Machine learning systems

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R

    1984-05-01

    With the dramatic rise of expert systems has come a renewed interest in the fuel that drives them-knowledge. For it is specialist knowledge which gives expert systems their power. But extracting knowledge from human experts in symbolic form has proved arduous and labour-intensive. So the idea of machine learning is enjoying a renaissance. Machine learning is any automatic improvement in the performance of a computer system over time, as a result of experience. Thus a learning algorithm seeks to do one or more of the following: cover a wider range of problems, deliver more accurate solutions, obtain answers more cheaply, and simplify codified knowledge. 6 references.

  16. Transfer Area Mechanical Handling Calculation

    International Nuclear Information System (INIS)

    Dianda, B.

    2004-01-01

    This calculation is intended to support the License Application (LA) submittal of December 2004, in accordance with the directive given by DOE correspondence received on the 27th of January 2004 entitled: ''Authorization for Bechtel SAX Company L.L. C. to Include a Bare Fuel Handling Facility and Increased Aging Capacity in the License Application, Contract Number DE-AC--28-01R W12101'' (Arthur, W.J., I11 2004). This correspondence was appended by further Correspondence received on the 19th of February 2004 entitled: ''Technical Direction to Bechtel SAIC Company L.L. C. for Surface Facility Improvements, Contract Number DE-AC--28-OIRW12101; TDL No. 04-024'' (BSC 2004a). These documents give the authorization for a Fuel Handling Facility to be included in the baseline. The purpose of this calculation is to establish preliminary bounding equipment envelopes and weights for the Fuel Handling Facility (FHF) transfer areas equipment. This calculation provides preliminary information only to support development of facility layouts and preliminary load calculations. The limitations of this preliminary calculation lie within the assumptions of section 5 , as this calculation is part of an evolutionary design process. It is intended that this calculation is superseded as the design advances to reflect information necessary to support License Application. The design choices outlined within this calculation represent a demonstration of feasibility and may or may not be included in the completed design. This calculation provides preliminary weight, dimensional envelope, and equipment position in building for the purposes of defining interface variables. This calculation identifies and sizes major equipment and assemblies that dictate overall equipment dimensions and facility interfaces. Sizing of components is based on the selection of commercially available products, where applicable. This is not a specific recommendation for the future use of these components or their

  17. 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.

  18. Remote handling equipment

    International Nuclear Information System (INIS)

    Clement, G.

    1984-01-01

    After a definition of intervention, problems encountered for working in an adverse environment are briefly analyzed for development of various remote handling equipments. Some examples of existing equipments are given [fr

  19. Ergonomics and patient handling.

    Science.gov (United States)

    McCoskey, Kelsey L

    2007-11-01

    This study aimed to describe patient-handling demands in inpatient units during a 24-hour period at a military health care facility. A 1-day total population survey described the diverse nature and impact of patient-handling tasks relative to a variety of nursing care units, patient characteristics, and transfer equipment. Productivity baselines were established based on patient dependency, physical exertion, type of transfer, and time spent performing the transfer. Descriptions of the physiological effect of transfers on staff based on patient, transfer, and staff characteristics were developed. Nursing staff response to surveys demonstrated how patient-handling demands are impacted by the staff's physical exertion and level of patient dependency. The findings of this study describe the types of transfers occurring in these inpatient units and the physical exertion and time requirements for these transfers. This description may guide selection of the most appropriate and cost-effective patient-handling equipment required for specific units and patients.

  20. Equipment system for advanced nuclear fuel development

    International Nuclear Information System (INIS)

    Kwon, Hyuk Il; Ji, C. G.; Bae, S. O.

    2002-11-01

    The purpose of the settlement of equipment system for nuclear Fuel Technology Development Facility(FTDF) is to build a seismic designed facility that can accommodate handling of nuclear materials including <20% enriched Uranium and produce HANARO fuel commercially, and also to establish the advanced common research equipment essential for the research on advanced fuel development. For this purpose, this research works were performed for the settlement of radiation protection system and facility special equipment for the FTDF, and the advanced common research equipment for the fuel fabrication and research. As a result, 11 kinds of radiation protection systems such as criticality detection and alarm system, 5 kinds of facility special equipment such as environmental pollution protection system and 5 kinds of common research equipment such as electron-beam welding machine were established. By the settlement of exclusive domestic facility for the research of advanced fuel, the fabrication and supply of HANARO fuel is possible and also can export KAERI-invented centrifugal dispersion fuel materials and its technology to the nations having research reactors in operation. For the future, the utilization of the facility will be expanded to universities, industries and other research institutes

  1. Automatic spent fuel ID number reader (I)

    International Nuclear Information System (INIS)

    Tanabe, S.; Kawamoto, H.; Fujimaki, K.; Kobe, A.

    1991-01-01

    An effective and efficient technique has been developed for facilitating identification works of LWR spent fuel stored in large scale spent fuel storage pools of such as processing plants. Experience shows that there are often difficulties in the implementation of operator's nuclear material accountancy and control works as well as safeguards inspections conducted on spent fuel assemblies stored in deep water pool. This paper reports that the technique is realized as an automatic spent fuel ID number reader system installed on fuel handling machine. The ID number reader system consists of an optical sub-system and an image processing sub-system. Thousands of spent fuel assemblies stored in under water open racks in each storage pool could be identified within relatively short time (e.g. within several hours) by using this combination. Various performance tests were carried out on image processing sub-system in 1990 using TV images obtained from different types of spent fuel assemblies stored in various storage pools of PWR and BWR power stations

  2. Ultrasonic decontamination of nuclear fuel. Feasibility study

    International Nuclear Information System (INIS)

    Berg, A.; Libal, A.; Norbaeck, J.; Wegemar, B.

    1995-05-01

    Ultrasonic decontamination of nuclear fuel is an expeditious way to reduce radiation exposures resulting in a minimal volume of waste. The fuel assemblies are set up in the fuel preparation machine one at a time and treated without prior disassemblage. By decontaminating 20% of the BWR fuel assemblies annually, there is a potential to reduce the collective dose by approximately 40-50%. Including also improved reactivity of the fuel, this amounts to an economic benefit of about 4 MSEK per reactor and year. The costs for performing the decontamination can be economically justified if the plants do not plan for short outages each year. The decontamination method could also be used for the purpose of removing tramp Uranium following a fuel failure or minor core accident. An additional benefit is removal of loosely adherent crud. The waste produced will be handled in a closed filtering circuit. The method is suggested to be verified in a test on discharged burnt-up fuel at site. The next step will be to develop the method further in order to be able to remove also tenacious crud. 12 refs, 4 tabs

  3. Evolution of on-power fuelling machines on Canadian natural uranium power reactors

    International Nuclear Information System (INIS)

    Isaac, P.

    1984-10-01

    The evolution of the on-power fuel changing process and fuelling machines on CANDU heavy-water pressure tube power reactors from the first nuclear power demonstration plant, 22 MWe NPD, to the latest plants now in design and development is described. The high availability of CANDU's is largely dependent on on-power fuelling. The on-power fuelling performance record of the 16 operating CANDU reactors, covering a 22 year period since the first plant became operational, is given. This shows that on-power fuel changing with light (unshielded), highly mobile and readily maintainable fuelling machines has been a success. The fuelling machines have contributed very little to the incapabilities of the plants and have been a key factor in placing CANDUs in the top ten list of world performance. Although fuel handling technology has reached a degree of maturity, refinements are continuing. A new single-ended fuel changing concept for horizontal reactors under development is described. This has the potential for reducing capital and operating costs for small reactors and increasing the fuelling capability of possible large reactors of the future

  4. Bulk handling benefits from ICT

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-11-15

    The efficiency and accuracy of bulk handling is being improved by the range of management information systems and services available today. As part of the program to extend Richards Bay Coal Terminal, Siemens is installing a manufacturing execution system which coordinates and monitors all movements of raw materials. The article also reports recent developments by AXSMarine, SunGuard Energy, Fuelworx and Railworx in providing integrated tools for tracking, managing and optimising solid/liquid fuels and rail car maintenance activities. QMASTOR Ltd. has secured a contract with Anglo Coal Australia to provide its Pit to Port.net{reg_sign} and iFuse{reg_sign} software systems across all their Australians sites, to include pit-to-product stockpile management. 2 figs.

  5. Simple machines

    CERN Document Server

    Graybill, George

    2007-01-01

    Just how simple are simple machines? With our ready-to-use resource, they are simple to teach and easy to learn! Chocked full of information and activities, we begin with a look at force, motion and work, and examples of simple machines in daily life are given. With this background, we move on to different kinds of simple machines including: Levers, Inclined Planes, Wedges, Screws, Pulleys, and Wheels and Axles. An exploration of some compound machines follows, such as the can opener. Our resource is a real time-saver as all the reading passages, student activities are provided. Presented in s

  6. SLSF loop handling system. Volume I. Structural analysis

    International Nuclear Information System (INIS)

    Ahmed, H.; Cowie, A.; Ma, D.

    1978-10-01

    SLSF loop handling system was analyzed for deadweight and postulated dynamic loading conditions, identified in Chapters II and III in Volume I of this report, using a linear elastic static equivalent method of stress analysis. Stress analysis of the loop handling machine is presented in Volume I of this report. Chapter VII in Volume I of this report is a contribution by EG and G Co., who performed the work under ANL supervision

  7. Hot Laboratories and Remote Handling

    International Nuclear Information System (INIS)

    2007-01-01

    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

  8. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi.

    1979-01-01

    Purpose: To prevent scattering of gaseous fission products released from fuel assemblies stored in an fbr type reactor. Constitution; A cap provided with means capable of storing gas is adapted to amount to the assembly handling head, for example, by way of threading in a storage rack of spent fuel assemblies consisting of a bottom plate, a top plate and an assembly support mechanism. By previously eliminating the gas inside of the assembly and the cap in the storage rack, gaseous fission products upon loading, if released from fuel rods during storage, are stored in the cap and do not scatter in the storage rack. (Horiuchi, T.)

  9. Development of tritium-handling technique

    International Nuclear Information System (INIS)

    Ohmura, Hiroshi; Hosaka, Akio; Okamoto, Takahumi

    1988-01-01

    The overview of developing activities for tritium-handling techniques in IHI are presented. To establish a fusion power plant, tritium handling is one of the key technologies. Recently in JAERI, conceptual design of FER (Fusion Experimental Reactor) has been carried out, and the FER system requires a processing system for a large amount of tritium. IHI concentrate on investigation of fuel gas purification, isotope separation and storage systems under contract with Toshiba Corporation. Design results of the systems and each components are reviewed. IHI has been developing fundamental handling techniques which are the ZrNi bed for hydrogen isotope storage and isotope separation by laser. The ZrNi bed with a tritium storage capacity of 1000 Ci has been constructed and recovery capability of the hydrogen isotope until 10 -4 Torr {0.013 Pa} was confirmed. In laser isotope separation, the optimum laser wave length has been determined. (author)

  10. MHSS: a material handling system simulator

    Energy Technology Data Exchange (ETDEWEB)

    Pomernacki, L.; Hollstien, R.B.

    1976-04-07

    A Material Handling System Simulator (MHSS) program is described that provides specialized functional blocks for modeling and simulation of nuclear material handling systems. Models of nuclear fuel fabrication plants may be built using functional blocks that simulate material receiving, storage, transport, inventory, processing, and shipping operations as well as the control and reporting tasks of operators or on-line computers. Blocks are also provided that allow the user to observe and gather statistical information on the dynamic behavior of simulated plants over single or replicated runs. Although it is currently being developed for the nuclear materials handling application, MHSS can be adapted to other industries in which material accountability is important. In this paper, emphasis is on the simulation methodology of the MHSS program with application to the nuclear material safeguards problem. (auth)

  11. Overview of HTGR fuel recycle

    International Nuclear Information System (INIS)

    Notz, K.J.

    1976-01-01

    An overview of HTGR fuel recycle is presented, with emphasis placed on reprocessing and fuel kernel refabrication. Overall recycle operations include (1) shipment and storage, (2) reprocessing, (3) refabrication, (4) waste handling, and (5) accountability and safeguards

  12. Practices of Handling

    DEFF Research Database (Denmark)

    Ræbild, Ulla

    to touch, pick up, carry, or feel with the hands. Figuratively it is to manage, deal with, direct, train, or control. Additionally, as a noun, a handle is something by which we grasp or open up something. Lastly, handle also has a Nordic root, here meaning to trade, bargain or deal. Together all four...... meanings seem to merge in the fashion design process, thus opening up for an embodied engagement with matter that entails direction giving, organizational management and negotiation. By seeing processes of handling as a key fashion methodological practice, it is possible to divert the discourse away from...... introduces four ways whereby fashion designers apply their own bodies as tools for design; a) re-activating past garment-design experiences, b) testing present garment-design experiences c) probing for new garment-design experiences and d) design of future garment experiences by body proxy. The paper...

  13. Remote handling at LAMPF

    International Nuclear Information System (INIS)

    Grisham, D.L.; Lambert, J.E.

    1983-01-01

    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

  14. 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

  15. Face machines

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, D.

    1999-06-01

    The article surveys latest equipment available from the world`s manufacturers of a range of machines for tunnelling. These are grouped under headings: excavators; impact hammers; road headers; and shields and tunnel boring machines. Products of thirty manufacturers are referred to. Addresses and fax numbers of companies are supplied. 5 tabs., 13 photos.

  16. Electric machine

    Science.gov (United States)

    El-Refaie, Ayman Mohamed Fawzi [Niskayuna, NY; Reddy, Patel Bhageerath [Madison, WI

    2012-07-17

    An interior permanent magnet electric machine is disclosed. The interior permanent magnet electric machine comprises a rotor comprising a plurality of radially placed magnets each having a proximal end and a distal end, wherein each magnet comprises a plurality of magnetic segments and at least one magnetic segment towards the distal end comprises a high resistivity magnetic material.

  17. Machine Learning.

    Science.gov (United States)

    Kirrane, Diane E.

    1990-01-01

    As scientists seek to develop machines that can "learn," that is, solve problems by imitating the human brain, a gold mine of information on the processes of human learning is being discovered, expert systems are being improved, and human-machine interactions are being enhanced. (SK)

  18. Nonplanar machines

    International Nuclear Information System (INIS)

    Ritson, D.

    1989-05-01

    This talk examines methods available to minimize, but never entirely eliminate, degradation of machine performance caused by terrain following. Breaking of planar machine symmetry for engineering convenience and/or monetary savings must be balanced against small performance degradation, and can only be decided on a case-by-case basis. 5 refs

  19. Advanced remote handling developments for high radiation applications

    International Nuclear Information System (INIS)

    Herndon, J.N.; Kring, C.T.; Feldman, M.J.; Kuban, D.P.; Martin, H.L.; Rowe, J.C.; Hamel, W.R.

    1985-01-01

    The Remote Control Engineering Task of the Consolidated Fuel Reprocessing Program at Oak Ridge National Laboratory has been developing advanced techniques for remote maintenance of future US fuel reprocessing plants. 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, and installation of a prototype remote handling system for reprocessing applications, the Advanced Integrated Maintenance System

  20. Equipment concepts for dry intercask transfer of spent fuel

    International Nuclear Information System (INIS)

    Schneider, K.J.

    1983-07-01

    This report documents the results of a study of preconceptual design and analysis of four intercask transfer concepts. The four concepts are: a large shielded cylindrical turntable that contains an integral fuel handling machine (turntable concept); a shielded fuel handling machine under which shipping and storage casks are moved horizontally (shuttle concept); a small hot cell containing equipment for transferring fuel between shipping and storage casks (that enter and leave the cell on carts) in a bifurcated trench (trench concept); and a large hot cell, shielded by an earthen berm, that houses equipment for handling fuel between casks that enter and leave the cell on a single cart (igloo concept). The casks considered in this study are most of the transport casks currently operable in the USA, and the storage casks designated REA-2023 and GNS Castor-V. Exclusive of basic services assumed to be provided at the host site, the design and capital costs are estimated to range from $9 to $13 million. The portion of capital costs for portable equipment (for potential later use at another site) was estimated to range from 70% to 98%, depending on the concept. Increasing portability from a range of 70 to 90% to 98% adds $2 to 4 million to the capital costs. Operating costs are estimated at about $2 million/year for all concepts. Implementation times range from about 18 months for the more conventional systems to 40 months for the more unique systems. Times and costs for relocation to another site are 10 to 14 months and about $1 million, plus shipping costs and costs of new construction at the new site. All concepts have estimated capacities for fuel transfer at least equal to the criterion set for this study. Only the hot cell concepts have capability for recanning or repair of canisters. Some development is believed to be required for the turntable and shuttle concepts, but none for the other two concepts

  1. Vestibule and Cask Preparation Mechanical Handling Calculation

    International Nuclear Information System (INIS)

    Ambre, N.

    2004-01-01

    The scope of this document is to develop the size, operational envelopes, and major requirements of the equipment to be used in the vestibule, cask preparation area, and the crane maintenance area of the Fuel Handling Facility. This calculation is intended to support the License Application (LA) submittal of December 2004, in accordance with the directive given by DOE correspondence received on the 27th of January 2004 entitled: ''Authorization for Bechtel SAIC Company L.L.C. to Include a Bare Fuel Handling Facility and Increased Aging Capacity in the License Application, Contract Number DE-AC--28-01R W12101'' (Ref. 167124). This correspondence was appended by further correspondence received on the 19th of February 2004 entitled: ''Technical Direction to Bechtel SAIC Company L.L. C. for Surface Facility Improvements, Contract Number DE-AC--28-01R W12101; TDL No. 04-024'' (Ref. 16875 1). These documents give the authorization for a Fuel Handling Facility to be included in the baseline. The limitations of this preliminary calculation lie within the assumptions of section 5 , as this calculation is part of an evolutionary design process

  2. Design Package for Fuel Retrieval System Fuel Handling Tool Modification

    International Nuclear Information System (INIS)

    TEDESCHI, D.J.

    2000-01-01

    This design package documents design, fabrication, and testing of new stinger tool design. Future revisions will document further development of the stinger tool and incorporate various developmental stages, and final test results

  3. Advanced SLARette delivery machine

    International Nuclear Information System (INIS)

    Bodner, R.R.

    1995-01-01

    SLARette 1 equipment, comprising of a SLARette Delivery Machine, SLAR Tools, SLAR power supplies and SLAR Inspection Systems was designed, developed and manufactured to service fuel channels of CANDU 6 stations during the regular yearly station outages. The Mark 2 SLARette Delivery Machine uses a Push Tube system to provide the axial and rotary movements of the SLAR Tool. The Push Tubes are operated remotely but must be attached and removed manually. Since this operation is performed at the Reactor face, there is radiation dose involved for the workers. An Advanced SLARette Delivery Machine which incorporates a computer controlled telescoping Ram in the place of the Push Tubes has been recently designed and manufactured. Utilization of the Advanced SLARette Delivery Machine significantly reduces the amount of radiation dose picked up by the workers because the need to have workers at the face of the Reactor during the SLARette operation is greatly reduced. This paper describes the design, development and manufacturing process utilized to produce the Advanced SLARette Delivery Machine and the experience gained during the Gentilly-2 NGS Spring outage. (author)

  4. Grain Grading and Handling.

    Science.gov (United States)

    Rendleman, Matt; Legacy, James

    This publication provides an introduction to grain grading and handling for adult students in vocational and technical education programs. Organized in five chapters, the booklet provides a brief overview of the jobs performed at a grain elevator and of the techniques used to grade grain. The first chapter introduces the grain industry and…

  5. Mars Sample Handling Functionality

    Science.gov (United States)

    Meyer, M. A.; Mattingly, R. L.

    2018-04-01

    The final leg of a Mars Sample Return campaign would be an entity that we have referred to as Mars Returned Sample Handling (MRSH.) This talk will address our current view of the functional requirements on MRSH, focused on the Sample Receiving Facility (SRF).

  6. Automation and remote handling activities in BARC: an overview

    International Nuclear Information System (INIS)

    Badodkar, D.N.

    2016-01-01

    Division of Remote Handling and Robotics, BARC has been working on design and development of various application specific remote handling and automation systems for nuclear front-end and back-end fuel cycle technologies. Division is also engaged in preservice and in-service inspection of coolant channels for Pressurized Heavy Water Reactors in India. Design and development of Reactor Control Mechanisms for Nuclear Research and Power Reactors (PHWRs and Compact LWRs) is another important activity carried out in this division. Robotic systems for Indoor and Outdoor surveillance in and around nuclear installations have also been developed. A line scan camera based system has been developed for measuring individual PHWR fuel pellet lengths as well as stack length. An industrial robot is used for autonomous exchange of pellets to achieve desired stack length. The system can be extended for active fuel pellets also. An automation system has been conceptualized for remote handling and transfer of spent fuel bundles from storage pool directly to the chopper unit of reprocessing plant. In case of Advanced Heavy Water Reactor which uses mixed oxides of (Th-Pu) and (Th-"2"3"3U ) as fuel, automation system for front-end fuel cycle has been designed, which includes Powder processing and pressing; Pellet handling and inspection; Pin handling and inspection; and Cluster assembly and dis-assembly in shielded facilities. System demonstration through fullscale mock-up facility is nearing completion. Above talk is presented on behalf of all the officers and staff of DRHR. The talk is mainly focused on development of an automated fuel fabrication facility for mixed oxides of (Th- Pu)/(Th-"2"3"3U ) fuel pins. An overview of divisional ongoing activities in the field of remote handling and automation are also covered. (author)

  7. Cellular Manufacturing System with Dynamic Lot Size Material Handling

    Science.gov (United States)

    Khannan, M. S. A.; Maruf, A.; Wangsaputra, R.; Sutrisno, S.; Wibawa, T.

    2016-02-01

    Material Handling take as important role in Cellular Manufacturing System (CMS) design. In several study at CMS design material handling was assumed per pieces or with constant lot size. In real industrial practice, lot size may change during rolling period to cope with demand changes. This study develops CMS Model with Dynamic Lot Size Material Handling. Integer Linear Programming is used to solve the problem. Objective function of this model is minimizing total expected cost consisting machinery depreciation cost, operating costs, inter-cell material handling cost, intra-cell material handling cost, machine relocation costs, setup costs, and production planning cost. This model determines optimum cell formation and optimum lot size. Numerical examples are elaborated in the paper to ilustrate the characterictic of the model.

  8. The Machine within the Machine

    CERN Multimedia

    Katarina Anthony

    2014-01-01

    Although Virtual Machines are widespread across CERN, you probably won't have heard of them unless you work for an experiment. Virtual machines - known as VMs - allow you to create a separate machine within your own, allowing you to run Linux on your Mac, or Windows on your Linux - whatever combination you need.   Using a CERN Virtual Machine, a Linux analysis software runs on a Macbook. When it comes to LHC data, one of the primary issues collaborations face is the diversity of computing environments among collaborators spread across the world. What if an institute cannot run the analysis software because they use different operating systems? "That's where the CernVM project comes in," says Gerardo Ganis, PH-SFT staff member and leader of the CernVM project. "We were able to respond to experimentalists' concerns by providing a virtual machine package that could be used to run experiment software. This way, no matter what hardware they have ...

  9. First neutrons from new machine

    International Nuclear Information System (INIS)

    Gray, D.A.

    1985-01-01

    Nimrod, the last weak focusing proton machine to be built, provided its first 7 GeV proton beam in 1963 and provided the research fuel for several generations of UK particle physicists. With the decision to build the SNS, the task was to transform the existing facility into a high repetition rate, high intensity machine furnishing the protons to bombard a neutron production target. As well as equipment from Nimrod, the SNS synchrotron also makes use of components from the old NINA electron machine at Daresbury, closed down in 1977. (orig./HSI).

  10. Machine translation

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, M

    1982-04-01

    Each language has its own structure. In translating one language into another one, language attributes and grammatical interpretation must be defined in an unambiguous form. In order to parse a sentence, it is necessary to recognize its structure. A so-called context-free grammar can help in this respect for machine translation and machine-aided translation. Problems to be solved in studying machine translation are taken up in the paper, which discusses subjects for semantics and for syntactic analysis and translation software. 14 references.

  11. Trends in remote handling device development

    International Nuclear Information System (INIS)

    Raimondi, T.

    1991-01-01

    A brief review is given of studies on layouts and methods for handling some major components requiring remote maintenance in future fusion reactors: Neutral sources and beam lines, the blanket, divertor plates, armour tiles and vacuum pumps. Comparison is made to problems encountered in JET, methods and equipment used and development work done there. Areas requiring development and research are outlined. These include topics which are the subject of papers presented here, such as dynamic studies and control of transporters, improvements to the man-machine interface and hot cell equipment. A variety of other topics where effort is needed are also mentioned: Environmental tolerance of components and equipment, TV viewing and compensation of viewing difficulties with aids such as computer graphics and image processing, safety assessment, computer aids for remote manipulation, remote cutting and welding techniques, routine in-vessel inspection methods and selection of connectors and flanges for remote handling. (orig.)

  12. Machine Learning

    CERN Multimedia

    CERN. Geneva

    2017-01-01

    Machine learning, which builds on ideas in computer science, statistics, and optimization, focuses on developing algorithms to identify patterns and regularities in data, and using these learned patterns to make predictions on new observations. Boosted by its industrial and commercial applications, the field of machine learning is quickly evolving and expanding. Recent advances have seen great success in the realms of computer vision, natural language processing, and broadly in data science. Many of these techniques have already been applied in particle physics, for instance for particle identification, detector monitoring, and the optimization of computer resources. Modern machine learning approaches, such as deep learning, are only just beginning to be applied to the analysis of High Energy Physics data to approach more and more complex problems. These classes will review the framework behind machine learning and discuss recent developments in the field.

  13. Machine Translation

    Indian Academy of Sciences (India)

    Research Mt System Example: The 'Janus' Translating Phone Project. The Janus ... based on laptops, and simultaneous translation of two speakers in a dialogue. For more ..... The current focus in MT research is on using machine learning.

  14. Gradient Boosting Machines, A Tutorial

    Directory of Open Access Journals (Sweden)

    Alexey eNatekin

    2013-12-01

    Full Text Available Gradient boosting machines are a family of powerful machine-learning techniques that have shown considerable success in a wide range of practical applications. They are highly customizable to the particular needs of the application, like being learned with respect to different loss functions. This article gives a tutorial introduction into the methodology of gradient boosting methods. A theoretical information is complemented with many descriptive examples and illustrations which cover all the stages of the gradient boosting model design. Considerations on handling the model complexity are discussed. A set of practical examples of gradient boosting applications are presented and comprehensively analyzed.

  15. Wastes from fuel reprocessing

    International Nuclear Information System (INIS)

    Eschrich, H.

    1976-01-01

    Handling, treatment, and interim storage of radioactive waste, problems confronted with during the reprocessing of spent fuel elements from LWR's according to the Purex-type process, are dealt with in detail. (HR/LN) [de

  16. Hot Laboratories and Remote Handling

    International Nuclear Information System (INIS)

    Bart, G.; Blanc, J.Y.; Duwe, R.

    2003-01-01

    The European Working Group on ' Hot Laboratories and Remote Handling' is firmly established as the major contact forum for the nuclear R and D facilities at the European scale. The yearly plenary meetings intend to: - Exchange experience on analytical methods, their implementation in hot cells, the methodologies used and their application in nuclear research; - Share experience on common infrastructure exploitation matters such as remote handling techniques, safety features, QA-certification, waste handling; - Promote normalization and co-operation, e.g., by looking at mutual complementarities; - Prospect present and future demands from the nuclear industry and to draw strategic conclusions regarding further needs. The 41. plenary meeting was held in CEA Saclay from September 22 to 24, 2003 in the premises and with the technical support of the INSTN (National Institute for Nuclear Science and Technology). The Nuclear Energy Division of CEA sponsored it. The Saclay meeting was divided in three topical oral sessions covering: - Post irradiation examination: new analysis methods and methodologies, small specimen technology, programmes and results; - Hot laboratory infrastructure: decommissioning, refurbishment, waste, safety, nuclear transports; - Prospective research on materials for future applications: innovative fuels (Generation IV, HTR, transmutation, ADS), spallation source materials, and candidate materials for fusion reactor. A poster session was opened to transport companies and laboratory suppliers. The meeting addressed in three sessions the following items: Session 1 - Post Irradiation Examinations. Out of 12 papers (including 1 poster) 7 dealt with surface and solid state micro analysis, another one with an equally complex wet chemical instrumental analytical technique, while the other four papers (including the poster) presented new concepts for digital x-ray image analysis; Session 2 - Hot laboratory infrastructure (including waste theme) which was

  17. Test sample handling apparatus

    International Nuclear Information System (INIS)

    1981-01-01

    A test sample handling apparatus using automatic scintillation counting for gamma detection, for use in such fields as radioimmunoassay, is described. The apparatus automatically and continuously counts large numbers of samples rapidly and efficiently by the simultaneous counting of two samples. By means of sequential ordering of non-sequential counting data, it is possible to obtain precisely ordered data while utilizing sample carrier holders having a minimum length. (U.K.)

  18. Handling and Transport Problems

    Energy Technology Data Exchange (ETDEWEB)

    Pomarola, J. [Head of Technical Section, Atomic Energy Commission, Saclay (France); Savouyaud, J. [Head of Electro-Mechanical Sub-Division, Atomic Energy Commission, Saclay (France)

    1960-07-01

    Arrangements for special or dangerous transport operations by road arising out of the activities of the Atomic Energy Commission are made by the Works and Installations Division which acts in concert with the Monitoring and Protection Division (MPD) whenever radioactive substances or appliances are involved. In view of the risk of irradiation and contamination entailed in handling and transporting radioactive substances, including waste, a specialized transport and storage team has been formed as a complement to the emergency and decontamination teams.

  19. Solid waste handling

    International Nuclear Information System (INIS)

    Parazin, R.J.

    1995-01-01

    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.)

  20. Handling of radioactive waste

    International Nuclear Information System (INIS)

    Sanhueza Mir, Azucena

    1998-01-01

    Based on characteristics and quantities of different types of radioactive waste produced in the country, achievements in infrastructure and the way to solve problems related with radioactive waste handling and management, are presented in this paper. Objectives of maintaining facilities and capacities for controlling, processing and storing radioactive waste in a conditioned form, are attained, within a great range of legal framework, so defined to contribute with safety to people and environment (au)

  1. Renal phosphate handling: Physiology

    Directory of Open Access Journals (Sweden)

    Narayan Prasad

    2013-01-01

    Full Text Available Phosphorus is a common anion. It plays an important role in energy generation. Renal phosphate handling is regulated by three organs parathyroid, kidney and bone through feedback loops. These counter regulatory loops also regulate intestinal absorption and thus maintain serum phosphorus concentration in physiologic range. The parathyroid hormone, vitamin D, Fibrogenic growth factor 23 (FGF23 and klotho coreceptor are the key regulators of phosphorus balance in body.

  2. Torus sector handling system

    International Nuclear Information System (INIS)

    Grisham, D.L.

    1981-01-01

    A remote handling system is proposed for moving a torus sector of the accelerator from under the cryostat to a point where it can be handled by a crane and for the reverse process for a new sector. Equipment recommendations are presented, as well as possible alignment schemes. Some general comments about future remote-handling methods and the present capabilities of existing systems will also be included. The specific task to be addressed is the removal and replacement of a 425 to 450 ton torus sector. This requires a horizontal movement of approx. 10 m from a normal operating position to a point where its further transport can be accomplished by more conventional means (crane or floor transporter). The same horizontal movement is required for reinstallation, but a positional tolerance of 2 cm is required to allow reasonable fit-up for the vacuum seal from the radial frames to the torus sector. Since the sectors are not only heavy but rather tall and narrow, the transport system must provide a safe, stable, and repeatable method fo sector movement. This limited study indicates that the LAMPF-based method of transporting torus sectors offers a proven method of moving heavy items. In addition, the present state of the art in remote equipment is adequate for FED maintenance

  3. Handling of Solid Residues

    International Nuclear Information System (INIS)

    Medina Bermudez, Clara Ines

    1999-01-01

    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

  4. Nuclear fuel production

    International Nuclear Information System (INIS)

    Randol, A.G.

    1985-01-01

    The production of new fuel for a power plant reactor and its disposition following discharge from the power plant is usually referred to as the ''nuclear fuel cycle.'' The processing of fuel is cyclic in nature since sometime during a power plant's operation old or ''depleted'' fuel must be removed and new fuel inserted. For light water reactors this step typically occurs once every 12-18 months. Since the time required for mining of the raw ore to recovery of reusable fuel materials from discharged materials can span up to 8 years, the management of fuel to assure continuous power plant operation requires simultaneous handling of various aspects of several fuel cycles, for example, material is being mined for fuel to be inserted in a power plant 2 years into the future at the same time fuel is being reprocessed from a discharge 5 years prior. Important aspects of each step in the fuel production process are discussed

  5. Enhanced Westinghouse WWER-1000 fuel design for Ukraine reactors

    International Nuclear Information System (INIS)

    Dye, M.; Shah, H.

    2015-01-01

    Westinghouse has completed design, development, and region quantity delivery of an enhanced Westinghouse fuel assembly for WWER-1000 reactors to support continued safe reactor operations. The enhanced design builds on the successful performance of an earlier generation design which has operated in the South Ukraine 3 reactor for multiple cycles without any fuel rod failures. Incorporated design enhancements include a thicker spacer grid outer strap, an enhanced spacer grid outer strap profile to limit the risk for, and impact of, mechanical interaction/interference with coresident fuel, an all Alloy 718 grid structure for improved stability and strength, and improvements to the top and bottom nozzles. Capable of meeting increased lateral loads generated from using a higher axial trip limit for the refueling machine crane, the design was verified by extensive mechanical and thermalhydraulic testing, which included a newly developed fuel assembly-to-fuel assembly handling test rig to assess performance during bounding core loading and unloading conditions. Through these extensive design enhancements and comprehensive testing program, the enhanced WWER-1000 design provides additional performance, handling, and reliability margins for safe reactor operation. (authors)

  6. Analysis of Virtual Sensors for Predicting Aircraft Fuel Consumption

    Data.gov (United States)

    National Aeronautics and Space Administration — Previous research described the use of machine learning algorithms to predict aircraft fuel consumption. This technique, known as Virtual Sensors, models fuel...

  7. Design guides for radioactive-material-handling facilities and equipment

    International Nuclear Information System (INIS)

    Doman, D.R.; Barker, R.E.

    1980-01-01

    Fourteen key areas relating to facilities and equipment for handling radioactive materials involved in examination, reprocessing, fusion fuel handling and remote maintenance have been defined and writing groups established to prepare design guides for each areas. The guides will give guidance applicable to design, construction, operation, maintenance and safety, together with examples and checklists. Each guide will be reviewed by an independent review group. The guides are expected to be compiled and published as a single document

  8. Preference Handling for Artificial Intelligence

    OpenAIRE

    Goldsmith, Judy; University of Kentucky; Junker, Ulrich; ILOG

    2009-01-01

    This article explains the benefits of preferences for AI systems and draws a picture of current AI research on preference handling. It thus provides an introduction to the topics covered by this special issue on preference handling.

  9. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    1983-01-01

    This report deals with certain aspects of the management of one of the most important wastes, i.e. the handling and storage of conditioned (immobilized and packaged) high-level waste from the reprocessing of spent nuclear fuel and, although much of the material presented here is based on information concerning high-level waste from reprocessing LWR fuel, the principles, as well as many of the details involved, are applicable to all fuel types. The report provides illustrative background material on the arising and characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The report introduces the principles important in conditioned high-level waste storage and describes the types of equipment and facilities, used or studied, for handling and storage of such waste. Finally, it discusses the safety and economic aspects that are considered in the design and operation of handling and storage facilities

  10. Refueling machine for a nuclear reactor

    International Nuclear Information System (INIS)

    Kowalski, E.F.; Hornak, L.P.; Swidwa, K.J.

    1981-01-01

    An improved refuelling machine for inserting and removing fuel assemblies from a nuclear reactor is described which has been designed to increase the reliability of such machines. The system incorporates features which enable the refuelling operation to be performed more efficiently and economically. (U.K.)

  11. Fuel element services

    International Nuclear Information System (INIS)

    Marta, H.; Alvarez, P.; Jimenez, J.

    2006-01-01

    Refuelling outages comprise a number of maintenance tasks scheduled long in advance to assure a reliable operation throughout the next cycle and, in the long run, a safer and more efficient plant. Most of these tasks are routine service of mechanical and electrical system and likewise fuel an be considered a critical component as to handling, inspection, cleaning and repair. ENUSA-ENWESA AIE has been working in this area since 1995 growing from fuel repair to a more integrated service that includes new and spent fuel handling, inserts, failed fuel rod detection systems, ultrasonic fuel cleaning, fuel repair and a comprehensive array of inspection and tests related to the reliability of the mechanical components in the fuel assembly, all this, performed in compliance with quality, safety, health physics and any other nuclear standard. (Author)

  12. Crud handling circuit

    International Nuclear Information System (INIS)

    Smith, J.C.; Manuel, R.J.; McAllister, J.E.

    1981-01-01

    A process for handling the problems of crud formation during the solvent extraction of wet-process phosphoric acid, e.g. for uranium and rare earth removal, is described. It involves clarification of the crud-solvent mixture, settling, water washing the residue and treatment of the crud with a caustic wash to remove and regenerate the solvent. Applicable to synergistic mixtures of dialkylphosphoric acids and trialkylphosphine oxides dissolved in inert diluents and more preferably to the reductive stripping technique. (U.K.)

  13. Handling of potassium

    International Nuclear Information System (INIS)

    Schwarz, N.; Komurka, M.

    1983-03-01

    As a result for the Fast Breeder Development extensive experience is available worldwide with respect to Sodium technology. Due to the extension of the research program to topping cycles with Potassium as the working medium, test facilities with Potassium have been designed and operated in the Institute of Reactor Safety. The different chemical properties of Sodium and Potassium give rise in new safety concepts and operating procedures. The handling problems of Potassium are described in the light of theoretical properties and own experiences. Selected literature on main safety and operating problems complete this report. (Author) [de

  14. Extreme coal handling

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, S; Homleid, D. [Air Control Science Inc. (United States)

    2004-04-01

    Within the journals 'Focus on O & M' is a short article describing modifications to coal handling systems at Eielson Air Force Base near Fairbanks, Alaska, which is supplied with power and heat from a subbituminous coal-fired central plant. Measures to reduce dust include addition of an enclosed recirculation chamber at each transfer point and new chute designs to reduce coal velocity, turbulence, and induced air. The modifications were developed by Air Control Science (ACS). 7 figs., 1 tab.

  15. Two-Dimensional Extreme Learning Machine

    Directory of Open Access Journals (Sweden)

    Bo Jia

    2015-01-01

    (BP networks. However, like many other methods, ELM is originally proposed to handle vector pattern while nonvector patterns in real applications need to be explored, such as image data. We propose the two-dimensional extreme learning machine (2DELM based on the very natural idea to deal with matrix data directly. Unlike original ELM which handles vectors, 2DELM take the matrices as input features without vectorization. Empirical studies on several real image datasets show the efficiency and effectiveness of the algorithm.

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Abe, Hideaki; Sakai, Takao; Ishida, Tomio; Yokota, Norikatsu.

    1992-01-01

    The lower ends of a plurality of plate-like shape memory alloys are secured at the periphery of the upper inside of the handling head of a fuel assembly. As the shape memory alloy, a Cu-Zn alloy, a Ti-Pd alloy or a Fe-Ni alloy is used. When high temperature coolants flow out to the handling head, the shape memory alloy deforms by warping to the outer side more greatly toward the upper portion thereof with the temperature increase of the coolants. As the result, the shape of the flow channel of the coolants is changed so as to enlarge at the exit of the upper end of the fuel assembly. Then, the pressure loss of the coolants in the fuel assembly is decreased by the enlargement. Accordingly, the flow rate of the coolants in the fuel assembly is increased to lower the temperature of the coolants. Further, high temperature coolants and low temperature coolants are mixed sufficiently just above the fuel assembly. This can suppress the temperature fluctuation of the mixed coolants in the upper portion of the reactor core, thereby enabling to decrease a fatigue and failures of the structural components in the upper portion of the reactor core. (I.N.)

  17. Machine Protection

    International Nuclear Information System (INIS)

    Zerlauth, Markus; Schmidt, Rüdiger; Wenninger, Jörg

    2012-01-01

    The present architecture of the machine protection system is being recalled and the performance of the associated systems during the 2011 run will be briefly summarized. An analysis of the causes of beam dumps as well as an assessment of the dependability of the machine protection systems (MPS) itself is being presented. Emphasis will be given to events that risked exposing parts of the machine to damage. Further improvements and mitigations of potential holes in the protection systems will be evaluated along with their impact on the 2012 run. The role of rMPP during the various operational phases (commissioning, intensity ramp up, MDs...) will be discussed along with a proposal for the intensity ramp up for the start of beam operation in 2012

  18. Machine Learning

    Energy Technology Data Exchange (ETDEWEB)

    Chikkagoudar, Satish; Chatterjee, Samrat; Thomas, Dennis G.; Carroll, Thomas E.; Muller, George

    2017-04-21

    The absence of a robust and unified theory of cyber dynamics presents challenges and opportunities for using machine learning based data-driven approaches to further the understanding of the behavior of such complex systems. Analysts can also use machine learning approaches to gain operational insights. In order to be operationally beneficial, cybersecurity machine learning based models need to have the ability to: (1) represent a real-world system, (2) infer system properties, and (3) learn and adapt based on expert knowledge and observations. Probabilistic models and Probabilistic graphical models provide these necessary properties and are further explored in this chapter. Bayesian Networks and Hidden Markov Models are introduced as an example of a widely used data driven classification/modeling strategy.

  19. Machine Protection

    CERN Document Server

    Zerlauth, Markus; Wenninger, Jörg

    2012-01-01

    The present architecture of the machine protection system is being recalled and the performance of the associated systems during the 2011 run will be briefly summarized. An analysis of the causes of beam dumps as well as an assessment of the dependability of the machine protection systems (MPS) itself is being presented. Emphasis will be given to events that risked exposing parts of the machine to damage. Further improvements and mitigations of potential holes in the protection systems will be evaluated along with their impact on the 2012 run. The role of rMPP during the various operational phases (commissioning, intensity ramp up, MDs...) will be discussed along with a proposal for the intensity ramp up for the start of beam operation in 2012.

  20. Machine Protection

    Energy Technology Data Exchange (ETDEWEB)

    Zerlauth, Markus; Schmidt, Rüdiger; Wenninger, Jörg [European Organization for Nuclear Research, Geneva (Switzerland)

    2012-07-01

    The present architecture of the machine protection system is being recalled and the performance of the associated systems during the 2011 run will be briefly summarized. An analysis of the causes of beam dumps as well as an assessment of the dependability of the machine protection systems (MPS) itself is being presented. Emphasis will be given to events that risked exposing parts of the machine to damage. Further improvements and mitigations of potential holes in the protection systems will be evaluated along with their impact on the 2012 run. The role of rMPP during the various operational phases (commissioning, intensity ramp up, MDs...) will be discussed along with a proposal for the intensity ramp up for the start of beam operation in 2012.

  1. Teletherapy machine

    International Nuclear Information System (INIS)

    Panyam, Vinatha S.; Rakshit, Sougata; Kulkarni, M.S.; Pradeepkumar, K.S.

    2017-01-01

    Radiation Standards Section (RSS), RSSD, BARC is the national metrology institute for ionizing radiation. RSS develops and maintains radiation standards for X-ray, beta, gamma and neutron radiations. In radiation dosimetry, traceability, accuracy and consistency of radiation measurements is very important especially in radiotherapy where the success of patient treatment is dependent on the accuracy of the dose delivered to the tumour. Cobalt teletherapy machines have been used in the treatment of cancer since the early 1950s and India had its first cobalt teletherapy machine installed at the Cancer Institute, Chennai in 1956

  2. Reactor fueling system

    International Nuclear Information System (INIS)

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

    Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

  3. Studies and research concerning BNFP: cask handling equipment standardization

    International Nuclear Information System (INIS)

    McCreery, P.N.

    1980-10-01

    This report covers the activities of one of the sub-tasks within the Spent LWR Fuel Transportation Receiving, Handling, and Storage program. The sub-task is identified as Cask Handling Equipment Standardization. The objective of the sub-task specifies: investigate and identify opportunities for standardization of cask interface equipment. This study will examine the potential benefits of standardized yokes, decontamination barriers and special tools, and, to the extent feasible, standardized methods and software for handling the variety of casks presently available in the US fleet. The result of the investigations is a compilation of reports that are related by their common goal of reducing cask turnaround time

  4. PRODUCTION OF NEW BIOMASS/WASTE-CONTAINING SOLID FUELS

    Energy Technology Data Exchange (ETDEWEB)

    David J. Akers; Glenn A. Shirey; Zalman Zitron; Charles Q. Maney

    2001-04-20

    CQ Inc. and its team members (ALSTOM Power Inc., Bliss Industries, McFadden Machine Company, and industry advisors from coal-burning utilities, equipment manufacturers, and the pellet fuels industry) addressed the objectives of the Department of Energy and industry to produce economical, new solid fuels from coal, biomass, and waste materials that reduce emissions from coal-fired boilers. This project builds on the team's commercial experience in composite fuels for energy production. The electric utility industry is interested in the use of biomass and wastes as fuel to reduce both emissions and fuel costs. In addition to these benefits, utilities also recognize the business advantage of consuming the waste byproducts of customers both to retain customers and to improve the public image of the industry. Unfortunately, biomass and waste byproducts can be troublesome fuels because of low bulk density, high moisture content, variable composition, handling and feeding problems, and inadequate information about combustion and emissions characteristics. Current methods of co-firing biomass and wastes either use a separate fuel receiving, storage, and boiler feed system, or mass burn the biomass by simply mixing it with coal on the storage pile. For biomass or biomass-containing composite fuels to be extensively used in the U.S., especially in the steam market, a lower cost method of producing these fuels must be developed that includes both moisture reduction and pelletization or agglomeration for necessary fuel density and ease of handling. Further, this method of fuel production must be applicable to a variety of combinations of biomass, wastes, and coal; economically competitive with current fuels; and provide environmental benefits compared with coal. Notable accomplishments from the work performed in Phase I of this project include the development of three standard fuel formulations from mixtures of coal fines, biomass, and waste materials that can be used in

  5. A Look at Technologies Vis-a-vis Information Handling Techniques.

    Science.gov (United States)

    Swanson, Rowena W.

    The paper examines several ideas for information handling implemented with new technologies that suggest directions for future development. These are grouped under the topic headings: Handling Large Data Banks, Providing Personalized Information Packages, Providing Information Specialist Services, and Expanding Man-Machine Interaction. Guides in…

  6. Automatic welding machine for piping

    International Nuclear Information System (INIS)

    Yoshida, Kazuhiro; Koyama, Takaichi; Iizuka, Tomio; Ito, Yoshitoshi; Takami, Katsumi.

    1978-01-01

    A remotely controlled automatic special welding machine for piping was developed. This machine is utilized for long distance pipe lines, chemical plants, thermal power generating plants and nuclear power plants effectively from the viewpoint of good quality control, reduction of labor and good controllability. The function of this welding machine is to inspect the shape and dimensions of edge preparation before welding work by the sense of touch, to detect the temperature of melt pool, inspect the bead form by the sense of touch, and check the welding state by ITV during welding work, and to grind the bead surface and inspect the weld metal by ultrasonic test automatically after welding work. The construction of this welding system, the main specification of the apparatus, the welding procedure in detail, the electrical source of this welding machine, the cooling system, the structure and handling of guide ring, the central control system and the operating characteristics are explained. The working procedure and the effect by using this welding machine, and the application to nuclear power plants and the other industrial field are outlined. The HIDIC 08 is used as the controlling computer. This welding machine is useful for welding SUS piping as well as carbon steel piping. (Nakai, Y.)

  7. Equipment to weld fuel rods of mixed oxides

    International Nuclear Information System (INIS)

    Aparicio, G.; Orlando, O.S.; Olano, V.R.; Toubes, B.; Munoz, C.A.

    1987-01-01

    Two welding outfits system T1G were designed and constructed to weld fuel rods with mixed oxides pellets (uranium and plutonium). One of them is connected to a glove box where the loading of sheaths takes place. The sheaths are driven to the welder through a removable plug pusher in the welding chamber. This equipment was designed to perform welding tests changing the parameters (gas composition and pressure, welding current, electrode position, etc.). The components of the welder, such as plug holder, chamber closure and peripheral accessories, were designed and constructed taking into account the working pressures in the machine, which is placed in a controlled area and connected to a glove box, where special safety conditions are necessary. The equipment to weld fuel bars is complemented by another machine, located in cold area, of the type presently used in the fuel elements factory. This equipment has been designed to perform some welding operations in sheaths and mixed oxide rods of the type Atucha I and II. Both machines have a programmed power supply of wide range and a vacuum, and pressurizing system that allows the change of parameters. Both systems have special features of handling and operation. (Author)

  8. BWR SFAT, gross-defect verification of spent BWR fuel. Final report on Task FIN A563 on the Finnish Support Programme to IAEA Safeguards including BWR SFAT User Manual

    International Nuclear Information System (INIS)

    Tarvainen, M.; Paakkunainen, M.; Tiitta, A.; Sarparanta, K.

    1994-04-01

    A measurement instrument called Spent Fuel Attribute Tester, SFAT, has been designed, fabricated and taken into use by the IAEA in gross defect verification of spent BWR fuel assemblies. The equipment consists of an underwater measurement head connected with cables to a control unit on the bridge of the fuel handling machine as well as to a PMCA for measurement of the gamma spectra. The BWR SFAT is optimized for the AFR interim storage, TVO KPA-STORE, of the TVO Power Company in Olkiluoto, Finland. It has a shape and it is moved like a fuel assembly using the fuel handling machine. No fuel movements are needed. Spent fuel specific radiation from the fission product 137 Cs at the gamma-ray energy of 662 keV is detected above the assemblies in the storage rack using a NaI(Tl) detector. In the design and in licensing the requirements of the IAEA, operator and the safety authority have been taken into account. The BWR SFAT allows modifications for other LWR fuel types with minor changes. The work has been carried out under the task FIN A 563 of the Finnish Support Programme to IAEA Safeguards. (orig.) (9 refs., 22 figs.)

  9. Machine testning

    DEFF Research Database (Denmark)

    De Chiffre, Leonardo

    This document is used in connection with a laboratory exercise of 3 hours duration as a part of the course GEOMETRICAL METROLOGY AND MACHINE TESTING. The exercise includes a series of tests carried out by the student on a conventional and a numerically controled lathe, respectively. This document...

  10. 25 CFR 547.11 - What are the minimum technical standards for money and credit handling?

    Science.gov (United States)

    2010-04-01

    ... GAMES § 547.11 What are the minimum technical standards for money and credit handling? This section... interface is: (i) Involved in the play of a game; (ii) In audit mode, recall mode or any test mode; (iii...) For machine-readable vouchers and coupons, a bar code or other form of machine readable representation...

  11. Machine rates for selected forest harvesting machines

    Science.gov (United States)

    R.W. Brinker; J. Kinard; Robert Rummer; B. Lanford

    2002-01-01

    Very little new literature has been published on the subject of machine rates and machine cost analysis since 1989 when the Alabama Agricultural Experiment Station Circular 296, Machine Rates for Selected Forest Harvesting Machines, was originally published. Many machines discussed in the original publication have undergone substantial changes in various aspects, not...

  12. Remote handling equipment for the decommissioning of the Windscale Advanced Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Barker, A.; Birss, I.R.; Fish, G.

    1984-01-01

    A decision to decommission the Windscale Advanced Gas Cooled Reactor was taken shortly after reactor shutdown in 1981. The fuel has now been discharged and the decommissioning programme will last about 10-12 years. The paper describes the programme and objectives and deals with methods of handling and disposing of the radioactive waste material. The main new facility required is a Waste Packaging Building adjacent to the existing reactor in which the waste boxes will be filled, active waste encapsulated in concrete and the boxes cleaned, swabbed and monitored to comply with IAEA transport regulations. The handling machine concept and features are described. The assaying and packaging of the waste material, the control of box movement and the process of concrete encapsulation is described. The paper concludes with a description of the development programme to support the Project. The tasks include a study of cutting techniques, production and control of dust and smoke, viewing and lighting methods, filtration, decontamination and fixing of contamination

  13. A dynamic simulation to study NET in-vessel handling operations

    International Nuclear Information System (INIS)

    Fung, P.T.F.

    1989-01-01

    The inspection, maintenance and repair of the Next European Torus (NET) fusion machine will require the extensive use of remote handling equipment to minimise the human exposure to the high radiation environment. The use of efficient manipulators will reduce the NET downtime by reducing the preparation time for entry into the controlled area and by performing the task with reasonable area and by performing the task with reasonable dexterity and speed, consistent with safety. A high fidelity simulation is a valuable tool to assist in the manipulator design, operations, trajectory planning, parameter optimisation and system verification. A manipulator simulation package called ASAD was originally developed by Spar for space manipulator applications. It is now being adapted to simulate the in-Vessel HandlingUnit for the NET program. This terestrial version of ASAD has been name ASAD - T. Spar, through the services of the Canadian Fusion Fuels Technology Project, is under contract to the NET program for the performance of this activity. This paper describes the capabillities and underlying assumptions of ASAD - T, aling with description of the simulation development of the NET in-vessel manipulator. (author). 4 refs.; 7 figs

  14. Remote handling in ZEPHYR

    International Nuclear Information System (INIS)

    Andelfinger, C.; Lackner, E.; Ulrich, M.; Weber, G.; Schilling, H.B.

    1982-04-01

    A conceptual design of the ZEPHYR building is described. The listed radiation data show that remote handling devices will be necessary in most areas of the building. For difficult repair and maintenance works it is intended to transfer complete units from the experimental hall to a hot cell which provides better working conditions. The necessary crane systems and other transport means are summarized as well as suitable commercially available manipulators and observation devices. The conept of automatic devices for cutting and welding and other operations inside the vacuum vessel and the belonging position control system is sketched. Guidelines for the design of passive components are set up in order to facilitate remote operation. (orig.)

  15. Handling hunger strikers.

    Science.gov (United States)

    1992-04-01

    Hunger strikes are being used increasingly and not only by those with a political point to make. Whereas in the past, hunger strikes in the United Kingdom seemed mainly to be started by terrorist prisoners for political purposes, the most recent was begun by a Tamil convicted of murder, to protest his innocence. In the later stages of his strike, before calling it off, he was looked after at the Hammersmith Hospital. So it is not only prison doctors who need to know how to handle a hunger strike. The following guidelines, adopted by the 43rd World Medical Assembly in Malta in November 1991, are therefore a timely reminder of the doctor's duties during a hunger strike.

  16. MFTF exception handling system

    International Nuclear Information System (INIS)

    Nowell, D.M.; Bridgeman, G.D.

    1979-01-01

    In the design of large experimental control systems, a major concern is ensuring that operators are quickly alerted to emergency or other exceptional conditions and that they are provided with sufficient information to respond adequately. This paper describes how the MFTF exception handling system satisfies these requirements. Conceptually exceptions are divided into one of two classes. Those which affect command status by producing an abort or suspend condition and those which fall into a softer notification category of report only or operator acknowledgement requirement. Additionally, an operator may choose to accept an exception condition as operational, or turn off monitoring for sensors determined to be malfunctioning. Control panels and displays used in operator response to exceptions are described

  17. Plutonium safe handling

    International Nuclear Information System (INIS)

    Tvehlov, Yu.

    2000-01-01

    The abstract, prepared on the basis of materials of the IAEA new leadership on the plutonium safe handling and its storage (the publication no. 9 in the Safety Reports Series), aimed at presenting internationally acknowledged criteria on the radiation danger evaluation and summarizing the experience in the safe management of great quantities of plutonium, accumulated in the nuclear states, is presented. The data on the weapon-class and civil plutonium, the degree of its danger, the measures for provision of its safety, including the data on accident radiation consequences with the fission number 10 18 , are presented. The recommendations, making it possible to eliminate the super- criticality danger, as well as ignition and explosion, to maintain the tightness of the facility, aimed at excluding the radioactive contamination and the possibility of internal irradiation, to provide for the plutonium security, physical protection and to reduce irradiation are given [ru

  18. Handle with care

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-03-15

    Full text: A film dealing with transport of radioactive materials by everyday means - rail, road, sea and air transport - has been made for IAEA. It illustrates in broad terms some of the simple precautions which should be followed by persons dealing with such materials during shipment. Throughout, the picture stresses the transport regulations drawn up and recommended by the Agency, and in particular the need to carry out carefully the instructions based on these regulations in order to ensure that there is no hazard to the public nor to those who handle radioactive materials in transit and storage. In straightforward language, the film addresses the porter of a goods wagon, an airline cargo clerk, a dockside crane operator, a truck driver and others who load and ship freight. It shows the various types of package used to contain different categories of radioactive substances according to the intensity of the radiation emitted. It also illustrates their robustness by a series of tests involving drops, fires, impact, crushing, etc. Clear instructions are conveyed on what to do in the event of an unlikely accident with any type of package. The film is entitled, 'The Safe Transport of Radioactive Materials', and is No. 3 in the series entitled, 'Handle with Care'. It was made for IAEA through the United Kingdom Atomic Energy Authority by the Film Producers' Guild in the United Kingdom. It is in 16 mm colour, optical sound, with a running time of 20 minutes. It is available for order at $50 either direct from IAEA or through any of its Member Governments. Prints can be supplied in English, French, Russian or Spanish. Copies are also available for adaptation for commentaries in other languages. (author)

  19. System for handling and storing radioactive waste

    Science.gov (United States)

    Anderson, John K.; Lindemann, Paul E.

    1984-01-01

    A system and method for handling and storing spent reactor fuel and other solid radioactive waste, including canisters to contain the elements of solid waste, storage racks to hold a plurality of such canisters, storage bays to store these racks in isolation by means of shielded doors in the bays. This system also includes means for remotely positioning the racks in the bays and an access tunnel within which the remotely operated means is located to position a rack in a selected bay. The modular type of these bays will facilitate the construction of additional bays and access tunnel extension.

  20. Remote handling equipment for CANDU retubing

    International Nuclear Information System (INIS)

    Crawford, G.S.; Lowe, H.

    1993-01-01

    Numet Engineering Ltd. has designed and supplied remote handling equipment for Ontario Hydro's retubing operation of its CANDU reactors at the Bruce Nuclear Generating Station. This equipment consists of ''Retubing Tool Carriers'' an'' Worktables'' which operate remotely or manually at the reactor face. Together they function to transport tooling to and from the reactor face, to position and support tooling during retubing operations, and to deliver and retrieve fuel channels and channel components. This paper presents the fundamentals of the process and discusses the equipment supplied in terms of its design, manufacturing, components and controls, to meet the functional and quality requirements of Ontario Hydro's retubing process. (author)