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Sample records for canister handling systems

  1. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    J.F. Beesley

    2005-04-21

    The purpose of this facility description document (FDD) is to establish requirements and associated bases that drive the design of the Canister Handling Facility (CHF), which will allow the design effort to proceed to license application. This FDD will be revised at strategic points as the design matures. This FDD identifies the requirements and describes the facility design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This FDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This FDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the facility. Knowledge of these requirements is essential in performing the design process. The FDD follows the design with regard to the description of the facility. The description provided in this FDD reflects the current results of the design process.

  2. CANISTER HANDLING FACILITY - VENTILATION CONFINEMENT ZONING ANALYSIS

    International Nuclear Information System (INIS)

    The purpose of this calculation is to calculate the necessary airflow distribution used to size the HVAC equipment for the Canister Handling Facility. These results will be compared to the Heating and Cooling Load Calculation in detailed design. The calculations contained in this document were developed by DandE/Mechanical HVAC and are intended solely for the use of the DandE/Mechanical HVAC department in its work regarding the HVAC system for the Canister Handling Facility. Yucca Mountain Project personnel from the DandE/Mechanical HVAC department should be consulted before use of the calculations for purposes other than those stated herein or used by individuals other than authorized personnel in DandE/Mechanical HVAC department

  3. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  4. CANISTER HANDLING FACILITY - VENTILATION AIR CALCULATION

    International Nuclear Information System (INIS)

    The purpose of this analysis is to establish the preliminary Ventilation Confinement Zone for the Canister Handling Facility (CHF). The results of this document will be used to determine the air quantities for each VCZ that will eventually be reflected in the development of the Ventilation Flow Diagrams. The analyses contained in this document are developed by D and E/Mechanical HVAC and are intended solely for the use of the D and E/Mechanical HVAC in its work regarding Confinement Zoning Analysis for the Canister Handling Facility. Yucca Mountain Project personnel from D and E/Mechanical HVAC should be consulted before use of the analyses for purposes other than those stated herein or used by individuals other than authorized personnel in D and E/Mechanical HVAC

  5. CANISTER HANDLING FACILITY WORKER DOSE ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    D.T. Dexheimer

    2004-02-27

    The purpose of this calculation is to estimate radiation doses received by personnel working in the Canister Handling Facility (CHF) performing operations to receive transportation casks, transfer wastes, prepare waste packages, perform associated equipment maintenance. The specific scope of work contained in this calculation covers individual worker group doses on an annual basis, and includes the contributions due to external and internal radiation. The results of this calculation will be used to support the design of the CHF and provide occupational dose estimates for the License Application.

  6. CANISTER HANDLING FACILITY WORKER DOSE ASSESSMENT

    International Nuclear Information System (INIS)

    The purpose of this calculation is to estimate radiation doses received by personnel working in the Canister Handling Facility (CHF) performing operations to receive transportation casks, transfer wastes, prepare waste packages, perform associated equipment maintenance. The specific scope of work contained in this calculation covers individual worker group doses on an annual basis, and includes the contributions due to external and internal radiation. The results of this calculation will be used to support the design of the CHF and provide occupational dose estimates for the License Application

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

    International Nuclear Information System (INIS)

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

  8. Groundwork for Universal Canister System Development

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Mike [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Prouty, Jeralyn L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Craig, Brian [Argonne National Lab. (ANL), Argonne, IL (United States); Han, Zenghu [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, John Hok [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Yung [Argonne National Lab. (ANL), Argonne, IL (United States); Pope, Ron [Argonne National Lab. (ANL), Argonne, IL (United States); Connolly, Kevin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feldman, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jarrell, Josh [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Radulescu, Georgeta [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wells, Alan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The mission of the United States Department of Energy's Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and go vernment - sponsored nuclear energy re search. S ome of the waste s that that must be managed have be en identified as good candidates for disposal in a deep borehole in crystalline rock (SNL 2014 a). In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister - based system that can be used for handling these wastes during the disposition process (i.e., storage, transfers, transportation, and disposal) could facilitate the eventual disposal of these wastes. This report provides information for a program plan for developing specifications regarding a canister - based system that facilitates small waste form packaging and disposal and that is integrated with the overall efforts of the DOE's Office of Nuclear Energy Used Fuel Dis position Camp aign's Deep Borehole Field Test . Groundwork for Universal Ca nister System Development September 2015 ii W astes to be considered as candidates for the universal canister system include capsules containing cesium and strontium currently stored in pools at the Hanford Site, cesium to be processed using elutable or nonelutable resins at the Hanford Site, and calcine waste from Idaho National Laboratory. The initial emphasis will be on disposal of the cesium and strontium capsules in a deep borehole that has been drilled into crystalline rock. Specifications for a universal canister system are derived from operational, performance, and regulatory requirements for storage, transfers, transportation, and disposal of radioactive waste. Agreements between the Department of Energy and the States of Washington and Idaho, as well as the Deep Borehole Field Test plan provide schedule requirements for development of the universal canister system

  9. Multi-purpose canister system evaluation: A systems engineering approach

    International Nuclear Information System (INIS)

    This report summarizes Department of Energy (DOE) efforts to investigate various container systems for handling, transporting, storing, and disposing of spent nuclear fuel (SNF) assemblies in the Civilian Radioactive Waste Management System (CRWMS). The primary goal of DOE's investigations was to select a container technology that could handle the vast majority of commercial SNF at a reasonable cost, while ensuring the safety of the public and protecting the environment. Several alternative cask and canister concepts were evaluated for SNF assembly packaging to determine the most suitable concept. Of these alternatives, the multi-purpose canister (MPC) system was determined to be the most suitable. Based on the results of these evaluations, the decision was made to proceed with design and certification of the MPC system. A decision to fabricate and deploy MPCs will be made after further studies and preparation of an environmental impact statement

  10. Shippingport Spent Fuel Canister System Description

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-03-27

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available.

  11. Shippingport Spent Fuel Canister System Description

    International Nuclear Information System (INIS)

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    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

  15. Analysis of Welding Joint on Handling High Level Waste-Glass Canister

    International Nuclear Information System (INIS)

    The analysis of welding joint of stainless steel austenitic AISI 304 for canister material has been studied. At the handling of waste-glass canister from melter below to interim storage, there is a step of welding of canister lid. Welding quality must be kept in a good condition, in order there is no gas out pass welding pores and canister be able to lift by crane. Two part of stainless steel plate in dimension (200 x 125 x 3) mm was jointed by welding. Welding was conducted by TIG machine with protection gas is argon. Electric current were conducted for welding were 70, 80, 90, 100, 110, 120, 130, and 140 A. Welded plates were cut with dimension according to JIS 3121 standard for tensile strength test. Hardness test in welding zone, HAZ, and plate were conducted by Vickers. Analysis of microstructure by optic microscope. The increasing of electric current at the welding, increasing of tensile strength of welding yields. The best quality welding yields using electric current was 110 A. At the welding with electric current more than 110 A, the electric current influence towards plate quality, so that decreasing of stainless steel plate quality and breaking at the plate. Tensile strength of stainless steel plate welding yields in requirement conditions according to application in canister transportation is 0.24 kg/mm2. (author)

  16. 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

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

    International Nuclear Information System (INIS)

    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

  18. Development of remote-operating welding system of canister cap

    International Nuclear Information System (INIS)

    The authors have developed a remote-operating welding system for mock-up test facilities of vitrification process of high level radio-active waste of nuclear fuel. This system enables cap sealing welding of canister to accomodate a vitrified waste. Supposing the operation is conducted under high level radio-active environment, the system has been considered to be well handled remotely by adopting guide-pin connection of the welding head, and also developing the automatic electrode exchanger, detecting method of work piece set location by means of the electrode itself as a sensor, slipping joint of power cable (work piece side) and shielding gas quality checking method by measuring an arc voltage changes. To ensure high quality welding, welding conditions were fully examined and established according to temperatures of the work piece before welding. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-03-09

    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

  20. Debris Removal Project K West Canister Cleaning System Performance Specification

    International Nuclear Information System (INIS)

    Approximately 2,300 metric tons Spent Nuclear Fuel (SNF) are currently stored within two water filled pools, the 105 K East (KE) fuel storage basin and the 105 K West (KW) fuel storage basin, at the U.S. Department of Energy, Richland Operations Office (RL). The SNF Project is responsible for operation of the K Basins and for the materials within them. A subproject to the SNF Project is the Debris Removal Subproject, which is responsible for removal of empty canisters and lids from the basins. Design criteria for a Canister Cleaning System to be installed in the KW Basin. This documents the requirements for design and installation of the system

  1. Debris Removal Project K West Canister Cleaning System Performance Specification

    Energy Technology Data Exchange (ETDEWEB)

    FARWICK, C.C.

    1999-12-09

    Approximately 2,300 metric tons Spent Nuclear Fuel (SNF) are currently stored within two water filled pools, the 105 K East (KE) fuel storage basin and the 105 K West (KW) fuel storage basin, at the U.S. Department of Energy, Richland Operations Office (RL). The SNF Project is responsible for operation of the K Basins and for the materials within them. A subproject to the SNF Project is the Debris Removal Subproject, which is responsible for removal of empty canisters and lids from the basins. Design criteria for a Canister Cleaning System to be installed in the KW Basin. This documents the requirements for design and installation of the system.

  2. 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)

    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

  3. Canister Cleaning System Final Design Report - Project A.2.A

    International Nuclear Information System (INIS)

    Approximately 2,300 metric tons Spent Nuclear Fuel (SNF) are currently stored within two water filled pools, the 105 K East (KE) fuel storage basin and the 105 K West (KW) fuel storage basin, at the U.S. Department of Energy, Richland Operations Office (RL). The SNF Project is responsible for operation of the K Basins and for the materials within them. A subproject to the SNF Project is the Debris Removal Subproject, which is responsible for removal of empty canisters and lids from the basins. The Canister Cleaning System (CCS) is part of the Debris Removal Project. The CCS will be installed in the KW Basin and operated during the fuel removal activity. The KW Basin has approximately 3600 canisters that require removal from the basin. The CCS is being designed to ''clean'' empty fuel canisters and lids and package them for disposal to the Environmental Restoration Disposal Facility complex. The system will interface with the KW Basin and be located in the Dummy Elevator Pit

  4. MFTF exception handling system

    International Nuclear Information System (INIS)

    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

  5. CLASSIFICATION OF THE MGR CANISTERED SNF DISPOSAL CONTAINER SYSTEM

    International Nuclear Information System (INIS)

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) canistered spent nuclear fuel disposal container system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333PY ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  6. Test report for the Sample Transfer Canister system

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, B.D.

    1998-03-04

    The Sample Transfer Canister will be used by the Waste Receiving and Processing Facility (WRAP) for the transport of small quantity liquid samples that meet the definition of a limited quantity radioactive material, and may also be corrosive and/or flammable. Transport of the system will typically be north of the Wye Barricade between WRAP and the 222-S Laboratory. The samples are intended to conform to the US Department of Transportation (DOT) regulation 49 CFR 1 73.4, ``Exceptions for small quantities.`` The regulations require prototype testing of the package to demonstrate the effectiveness of the packaging system. The test procedure consisted of one 24-hour compression test and five drop tests of various orientations onto an unyielding drop pad. The testing of the Sample Transfer Canister System was performed between February 16, 1998 and February 25, 1998. The results of the testing concluded that the Sample Transfer Canister System successfully met the testing requirements with certain modifications to the original system. The modifications included replacing the original eight flange screws which were cold rolled 316 stainless steel with greater strength grade 8 high carbon-carbon steel screws, replacing the initial two glass receptacles with a better performing single glass receptacle which proved not to leak during testing, and adding more bubble wrap as extra padding.

  7. Shielded canister transporter equipment acceptance test operations

    International Nuclear Information System (INIS)

    The defense waste processing facility (DWPF) processes high level waste at the Savannah River Plant (SRP) by vitrifying the waste and placing it in stainless stell canisters for long term storage. The shielded canister transporter (SCT) is a diesel powered mobile rubber tired self-propelled vehicle which transports the canisters from the DWPF processing facility to the on-site waste storage building. The SCT has a system of automatic programmable logic controls (PLC) which provides operational handling control with a shielded transfer cask and associated canister positional equipment

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-05-15

    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. Tritium handling in vacuum systems

    Energy Technology Data Exchange (ETDEWEB)

    Gill, J.T. [Monsanto Research Corp., Miamisburg, OH (United States). Mound Facility; Coffin, D.O. [Los Alamos National Lab., NM (United States)

    1986-10-01

    This report provides a course in Tritium handling in vacuum systems. Topics presented are: Properties of Tritium; Tritium compatibility of materials; Tritium-compatible vacuum equipment; and Tritium waste treatment.

  11. Lunar Materials Handling System Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Lunar Materials Handling System (LMHS) is a method for transfer of bulk materials and products into and out of process equipment in support of lunar and Mars in...

  12. Lunar Materials Handling System Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Lunar Materials Handling System (LMHS) is a method for transfer of lunar soil into and out of process equipment in support of in situ resource utilization...

  13. Commercial radioactive waste management system feasibility with the universal canister concept. Volume 1

    International Nuclear Information System (INIS)

    A Program Research and Development Announcement (PRDA) was initiated by DOE to solicit from industry new and novel ideas for improvements in the nuclear waste management system. GA Technologies Inc. was contracted to study a system utilizing a universal canister which could be loaded at the reactor and used throughout the waste management system. The proposed canister was developed with the objective of meeting the mission requirements with maximum flexibility and at minimum cost. Canister criteria were selected from a thorough analysis of the spent fuel inventory, and canister concepts were evaluated along with the shipping and storage casks to determine the maximum payload. Engineering analyses were performed on various cask/canister combinations. One important criterion was the interchangeability of the canisters between truck and rail cask systems. A canister was selected which could hold three PWR intact fuel elements or up to eight consolidated PWR fuel elements. One canister could be shipped in an overweight truck cask or six in a rail cask. Economic analysis showed a cost savings of the reference system under consideration at that time

  14. System design description for the consolidated sludge sampling system for K Basins floor and fuel canisters

    International Nuclear Information System (INIS)

    This System Design Description describes the Consolidated Sludge Sampling System used in the gathering of sludge samples from K Basin floor and fuel canisters. This document provides additional information on the need for the system, the functions and requirements of the systems, the operations of the system, and the general work plan used in its' design and development

  15. Retrievability of spent nuclear fuel canisters

    International Nuclear Information System (INIS)

    As a part of the designing process of the Finnish spent nuclear fuel repository, a preliminary study has been carried out to investigate how the canisters could technically be retrieved to the ground surface. Possibility of retrieving a canister has been investigated in different phases of the disposal project. Retrievability has not been a design goal for the spent fuel repository. However, design of the repository includes some features that may ease the retrieval of canisters in the future. Spent fuel elements are packaged in massive copper-iron canisters, which are mechanically strong and long-lived. The repository consists of excavated tunnels in hard rock which are supposed to be very long-lived making the removal of the tunnel backfilling technically possible also in the future. As long as the bentonite buffer has not been installed the canister can be returned to the ground surface using the same equipment as was used when the canister was brought down to the repository and lowered into the hole. In the encapsulation station the spent fuel elements can be packaged in the other canister or in the transport cask. After a deposition tunnel has been backfilled and closed, the retrieval consists of tearing down the concrete structure at the entry of the deposition tunnel, removal of the tunnel backfilling, removal of the bentonite from the disposal hole and lifting up of the canister. Various methods, e.g., flushing the bentonite with saline solutions, can be used to detach the canister from a hole with fully saturated bentonite. Recovery will be technically possible also after closing of the disposal facility. Backfilling of the shafts and tunnels will be removed and additional new structures and systems will have to be built in the repository. After that canisters can be transported to the ground surface as described above. In addition, handling of the canisters at the ground surface will require additional facilities. Canisters can be packaged in the

  16. Incidence Handling and Response System

    CERN Document Server

    Kalbande, Prof Dhananjay R; Singh, Mr Manish

    2009-01-01

    A computer network can be attacked in a number of ways. The security-related threats have become not only numerous but also diverse and they may also come in the form of blended attacks. It becomes difficult for any security system to block all types of attacks. This gives rise to the need of an incidence handling capability which is necessary for rapidly detecting incidents, minimizing loss and destruction, mitigating the weaknesses that were exploited and restoring the computing services. Incidence response has always been an important aspect of information security but it is often overlooked by security administrators. in this paper, we propose an automated system which will handle the security threats and make the computer network capable enough to withstand any kind of attack. we also present the state-of-the-art technology in computer, network and software which is required to build such a system.

  17. Inter-process handling automating system; Koteikan handling jidoka system

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, H. [Meidensha Corp., Tokyo (Japan)

    1994-10-18

    This paper introduces automation of loading works in production site by using robots. Loading robots are required of complex movements, and are used for loading work in processing machines requiring six degrees of freedom and for relatively simple palletizing work that can be dealt with by four degrees of freedom. The `inter-machine handling system` is an automated system performed by a ceiling running robot in which different workpiece model determination and positional shift measurement are carried out by image processing. A robot uses the image information to exchange hands automatically as required, and clamp a workpiece; then runs to an M/C to replace the processed workpiece; and put the M/C processes workpiece onto a multi-axial dedicated machine. Five processing machines are operated in parallel with the cycle time matched with that of this handling process, and a processing machine finished of processing is given a handling work in preferential order. As a result, improvement in productivity and elimination of two workers were achieved simultaneously. 6 figs., 5 tabs.

  18. SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB) MULTI CANISTER OVERPACK (MCO) SAMPLING SYSTEM VALIDATION (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    BLACK, D.M.; KLEM, M.J.

    2003-11-17

    Approximately 400 Multi-canister overpacks (MCO) containing spent nuclear fuel are to be interim stored at the Canister Storage Building (CSB). Several MCOs (monitored MCOs) are designated to be gas sampled periodically at the CSB sampling/weld station (Bader 2002a). The monitoring program includes pressure, temperature and gas composition measurements of monitored MCOs during their first two years of interim storage at the CSB. The MCO sample cart (CART-001) is used at the sampling/weld station to measure the monitored MCO gas temperature and pressure, obtain gas samples for laboratory analysis and refill the monitored MCO with high purity helium as needed. The sample cart and support equipment were functionally and operationally tested and validated before sampling of the first monitored MCO (H-036). This report documents the results of validation testing using training MCO (TR-003) at the CSB. Another report (Bader 2002b) documents the sample results from gas sampling of the first monitored MCO (H-036). Validation testing of the MCO gas sampling system showed the equipment and procedure as originally constituted will satisfactorily sample the first monitored MCO. Subsequent system and procedural improvements will provide increased flexibility and reliability for future MCO gas sampling. The physical operation of the sampling equipment during testing provided evidence that theoretical correlation factors for extrapolating MCO gas composition from sample results are unnecessarily conservative. Empirically derived correlation factors showed adequate conservatism and support use of the sample system for ongoing monitored MCO sampling.

  19. Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

    International Nuclear Information System (INIS)

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

  20. Incidence Handling and Response System

    OpenAIRE

    Kalbande, Prof. Dhananjay R.; Thampi, Dr. G. T.; Singh, Manish

    2009-01-01

    A computer network can be attacked in a number of ways. The security-related threats have become not only numerous but also diverse and they may also come in the form of blended attacks. It becomes difficult for any security system to block all types of attacks. This gives rise to the need of an incidence handling capability which is necessary for rapidly detecting incidents, minimizing loss and destruction, mitigating the weaknesses that were exploited and restoring the computing services. I...

  1. A Mars Sample Return Sample Handling System

    Science.gov (United States)

    Wilson, David; Stroker, Carol

    2013-01-01

    We present a sample handling system, a subsystem of the proposed Dragon landed Mars Sample Return (MSR) mission [1], that can return to Earth orbit a significant mass of frozen Mars samples potentially consisting of: rock cores, subsurface drilled rock and ice cuttings, pebble sized rocks, and soil scoops. The sample collection, storage, retrieval and packaging assumptions and concepts in this study are applicable for the NASA's MPPG MSR mission architecture options [2]. Our study assumes a predecessor rover mission collects samples for return to Earth to address questions on: past life, climate change, water history, age dating, understanding Mars interior evolution [3], and, human safety and in-situ resource utilization. Hence the rover will have "integrated priorities for rock sampling" [3] that cover collection of subaqueous or hydrothermal sediments, low-temperature fluidaltered rocks, unaltered igneous rocks, regolith and atmosphere samples. Samples could include: drilled rock cores, alluvial and fluvial deposits, subsurface ice and soils, clays, sulfates, salts including perchlorates, aeolian deposits, and concretions. Thus samples will have a broad range of bulk densities, and require for Earth based analysis where practical: in-situ characterization, management of degradation such as perchlorate deliquescence and volatile release, and contamination management. We propose to adopt a sample container with a set of cups each with a sample from a specific location. We considered two sample cups sizes: (1) a small cup sized for samples matching those submitted to in-situ characterization instruments, and, (2) a larger cup for 100 mm rock cores [4] and pebble sized rocks, thus providing diverse samples and optimizing the MSR sample mass payload fraction for a given payload volume. We minimize sample degradation by keeping them frozen in the MSR payload sample canister using Peltier chip cooling. The cups are sealed by interference fitted heat activated memory

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

    International Nuclear Information System (INIS)

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

  3. Remote Handling System for Ignitor^*

    Science.gov (United States)

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

    2005-10-01

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

  4. Lignite and conditioned ash handling systems

    Energy Technology Data Exchange (ETDEWEB)

    Bibolini, P.; Di Giacomo, L.; Ruga, A.M. [Techint (Italy)

    2001-10-01

    This article discusses Techint's latest contract for the engineering and supply of a lignite and conditioned ash handling system. Techint Italimpianti, the materials handling unit of Techint Technologies has served the market for over 40 years as a leading supplier of a range of systems for the handling of iron ore, pellets, coal, cement, bauxite, and aluminium. 6 figs.

  5. System Configuration Management Implementation Procedure for the Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    GARRISON, R.C.

    2000-11-28

    This document provides configuration management for the Distributed Control System (DCS), the Gaseous Effluent Monitoring System (GEMS-100) System, the Heating Ventilation and Air Conditioning (HVAC) Programmable Logic Controller (PLC), the Canister Receiving Crane (CRC) CRN-001 PLC, and both North and South vestibule door interlock system PLCs at the Canister Storage Building (CSB). This procedure identifies and defines software configuration items in the CSB control and monitoring systems, and defines configuration control throughout the system life cycle. Components of this control include: configuration status accounting; physical protection and control; and verification of the completeness and correctness of these items.

  6. ATA diagnostic data handling system: an overview

    International Nuclear Information System (INIS)

    The functions to be performed by the ATA diagnostic data handling system are discussed. The capabilities of the present data acquisition system (System 0) are presented. The goals for the next generation acquisition system (System 1), currently under design, are discussed. Facilities on the Octopus system for data handling are reviewed. Finally, we discuss what has been learned about diagnostics and computer based data handling during the past year

  7. Temperature history for canistered fuel lag storage areas during the loss of cooling air at the receiving and handling building of the MRS Facility

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory has analyzed the temperature history at a canistered fuel lag storage area during a postulated failure of the cooling system. A two-dimensional analysis was performed using the GE2D computer code, which accounts for thermal radiation, conduction, and convective heat transfer. The results indicate that the system may not reach a steady-state condition because heat transfer through the top and the bottom of the system is not enough to remove the energy generated in the canistered fuel. Although limits for abnormal operation have not been set, the temperatures do not reach limiting conditions for normal operation for 32 hours, which should be enough time to repair the cooling system

  8. An evaluation of dual-purpose canisters in the Civilian Radioactive Waste Management System

    International Nuclear Information System (INIS)

    An evaluation was made of the Civilian Radioactive Waste Management System (CRWMS) using dual-purpose canisters (DPCs) and was compared to a system using multi-purpose canisters (MPCs). The DPC would be designed for transportation and storage, whereas the MPC is designed for transportation, storage, and geologic disposal. Implementation of the DPC concept could allow the federal government to proceed with storage and transportation of spent nuclear fuel (SNF) without linkage to geologic disposal, while continuing to independently develop ultimate geologic disposal requirements and designs

  9. Cask system design guidance for robotic handling

    International Nuclear Information System (INIS)

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

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  13. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    International Nuclear Information System (INIS)

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  14. System-Level Logistics for Dual Purpose Canister Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kalinina, Elena A.

    2014-06-03

    The analysis presented in this report investigated how the direct disposal of dual purpose canisters (DPCs) may be affected by the use of standard transportation aging and disposal canisters (STADs), early or late start of the repository, and the repository emplacement thermal power limits. The impacts were evaluated with regard to the availability of the DPCs for emplacement, achievable repository acceptance rates, additional storage required at an interim storage facility (ISF) and additional emplacement time compared to the corresponding repackaging scenarios, and fuel age at emplacement. The result of this analysis demonstrated that the biggest difference in the availability of UNF for emplacement between the DPC-only loading scenario and the DPCs and STADs loading scenario is for a repository start date of 2036 with a 6 kW thermal power limit. The differences are also seen in the availability of UNF for emplacement between the DPC-only loading scenario and the DPCs and STADs loading scenario for the alternative with a 6 kW thermal limit and a 2048 start date, and for the alternatives with a 10 kW thermal limit and 2036 and 2048 start dates. The alternatives with disposal of UNF in both DPCs and STADs did not require additional storage, regardless of the repository acceptance rate, as compared to the reference repackaging case. In comparison to the reference repackaging case, alternatives with the 18 kW emplacement thermal limit required little to no additional emplacement time, regardless of the repository start time, the fuel loading scenario, or the repository acceptance rate. Alternatives with the 10 kW emplacement thermal limit and the DPCs and STADs fuel loading scenario required some additional emplacement time. The most significant decrease in additional emplacement time occurred in the alternative with the 6 kW thermal limit and the 2036 repository starting date. The average fuel age at emplacement ranges from 46 to 88 years. The maximum fuel age at

  15. Evaluation of DUSTRAN Software System for Modeling Chloride Deposition on Steel Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Tracy T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fritz, Brad G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rutz, Frederick C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Devanathan, Ram [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-07-29

    The degradation of steel by stress corrosion cracking (SCC) when exposed to atmospheric conditions for decades is a significant challenge in the fossil fuel and nuclear industries. SCC can occur when corrosive contaminants such as chlorides are deposited on a susceptible material in a tensile stress state. The Nuclear Regulatory Commission has identified chloride-induced SCC as a potential cause for concern in stainless steel used nuclear fuel (UNF) canisters in dry storage. The modeling of contaminant deposition is the first step in predictive multiscale modeling of SCC that is essential to develop mitigation strategies, prioritize inspection, and ensure the integrity and performance of canisters, pipelines, and structural materials. A multiscale simulation approach can be developed to determine the likelihood that a canister would undergo SCC in a certain period of time. This study investigates the potential of DUSTRAN, a dust dispersion modeling system developed by Pacific Northwest National Laboratory, to model the deposition of chloride contaminants from sea salt aerosols on a steel canister. Results from DUSTRAN simulations run with historical meteorological data were compared against measured chloride data at a coastal site in Maine. DUSTRAN’s CALPUFF model tended to simulate concentrations higher than those measured; however, the closest estimations were within the same order of magnitude as the measured values. The decrease in discrepancies between measured and simulated values as the level of abstraction in wind speed decreased suggest that the model is very sensitive to wind speed. However, the influence of other parameters such as the distinction between open-ocean and surf-zone sources needs to be explored further. Deposition values predicted by the DUSTRAN system were not in agreement with concentration values and suggest that the deposition calculations may not fully represent physical processes. Overall, results indicate that with parameter

  16. Evaluation of DUSTRAN Software System for Modeling Chloride Deposition on Steel Canisters

    International Nuclear Information System (INIS)

    The degradation of steel by stress corrosion cracking (SCC) when exposed to atmospheric conditions for decades is a significant challenge in the fossil fuel and nuclear industries. SCC can occur when corrosive contaminants such as chlorides are deposited on a susceptible material in a tensile stress state. The Nuclear Regulatory Commission has identified chloride-induced SCC as a potential cause for concern in stainless steel used nuclear fuel (UNF) canisters in dry storage. The modeling of contaminant deposition is the first step in predictive multiscale modeling of SCC that is essential to develop mitigation strategies, prioritize inspection, and ensure the integrity and performance of canisters, pipelines, and structural materials. A multiscale simulation approach can be developed to determine the likelihood that a canister would undergo SCC in a certain period of time. This study investigates the potential of DUSTRAN, a dust dispersion modeling system developed by Pacific Northwest National Laboratory, to model the deposition of chloride contaminants from sea salt aerosols on a steel canister. Results from DUSTRAN simulations run with historical meteorological data were compared against measured chloride data at a coastal site in Maine. DUSTRAN's CALPUFF model tended to simulate concentrations higher than those measured; however, the closest estimations were within the same order of magnitude as the measured values. The decrease in discrepancies between measured and simulated values as the level of abstraction in wind speed decreased suggest that the model is very sensitive to wind speed. However, the influence of other parameters such as the distinction between open-ocean and surf-zone sources needs to be explored further. Deposition values predicted by the DUSTRAN system were not in agreement with concentration values and suggest that the deposition calculations may not fully represent physical processes. Overall, results indicate that with parameter

  17. Thermal analysis of dry concrete canister storage system for CANDU spent fuel

    International Nuclear Information System (INIS)

    This paper presents the results of a thermal analysis of the concrete canisters for interim dry storage of spent, irradiated Canadian Deuterium Uranium(CANDU) fuel. The canisters are designed to contain 6-year-old fuel safely for periods of 50 years in stainless steel baskets sealed inside a steel-lined concrete shield. In order to assure fuel integrity during the storage, fuel rod temperature shall not exceed the temperature limit. The contents of thermal analysis include the following : 1) Steady state temperature distributions under the conservative ambient temperature and insolation load. 2) Transient temperature distributions under the changes in ambient temperature and insolation load. Accounting for the coupled heat transfer modes of conduction, convection, and radiation, the computer code HEATING5 was used to predict the thermal response of the canister storage system. As HEATING5 does not have the modeling capability to compute radiation heat transfer on a rod-to-rod basis, a separate calculating routine was developed and applied to predict temperature distribution in a fuel bundle. Thermal behavior of the canister is characterized by the large thermal mass of the concrete and radiative heat transfer within the basket. The calculated results for the worst case (steady state with maximum ambient temperature and design insolation load) indicated that the maximum temperature of the 6 year cooled fuel reached to 182.4 .deg. C, slightly above the temperature limit of 180 .deg. C. However,the thermal inertia of the thick concrete wall moderates the internal changes and prevents a rise in fuel temperature in response to ambient changes. The maximum extent of the transient zone was less than 75% of the concrete wall thickness for cyclic insolation changes. When transient nature of ambient temperature and insolation load are considered, the fuel temperature will be a function of the long term ambient temperature as opposed to daily extremes. The worst design

  18. GeoLab Sample Handling System Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Develop  a robotic sample handling/ manipulator system for the GeoLab glovebox. This work leverages from earlier GeoLab work and a 2012 collaboration with a...

  19. WASTE HANDLING BUILDING VENTILATION SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    P.A. Kumar

    2000-06-21

    The Waste Handling Building Ventilation System provides heating, ventilation, and air conditioning (HVAC) for the contaminated, potentially contaminated, and uncontaminated areas of the Monitored Geologic Repository's (MGR) Waste Handling Building (WHB). In the uncontaminated areas, the non-confinement area ventilation system maintains the proper environmental conditions for equipment operation and personnel comfort. In the contaminated and potentially contaminated areas, in addition to maintaining the proper environmental conditions for equipment operation and personnel comfort, the contamination confinement area ventilation system directs potentially contaminated air away from personnel in the WHB and confines the contamination within high-efficiency particulate air (HEPA) filtration units. The contamination confinement areas ventilation system creates airflow paths and pressure zones to minimize the potential for spreading contamination within the building. The contamination confinement ventilation system also protects the environment and the public by limiting airborne releases of radioactive or other hazardous contaminants from the WHB. The Waste Handling Building Ventilation System is designed to perform its safety functions under accident conditions and other Design Basis Events (DBEs) (such as earthquakes, tornadoes, fires, and loss of the primary electric power). Additional system design features (such as compartmentalization with independent subsystems) limit the potential for cross-contamination within the WHB. The system provides status of important system parameters and equipment operation, and provides audible and/or visual indication of off-normal conditions and equipment failures. The Waste Handling Building Ventilation System confines the radioactive and hazardous material within the building such that the release rates comply with regulatory limits. The system design, operations, and maintenance activities incorporate ALARA (as low as is

  20. Automated system for handling tritiated mixed waste

    International Nuclear Information System (INIS)

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

  1. Acceptance of canisters of consolidated spent nuclear fuel by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for deliver; and defense and commercial high-level waste packages. This document discusses canister standards and criteria. 12 refs., 7 figs., 28 tabs

  2. Acceptance of canisters of consolidated spent nuclear fuel by the Federal Waste Management System

    Energy Technology Data Exchange (ETDEWEB)

    1990-03-01

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for deliver; and defense and commercial high-level waste packages. This document discusses canister standards and criteria. 12 refs., 7 figs., 28 tabs.

  3. Thermal analysis of heat storage canisters for a solar dynamic, space power system

    Science.gov (United States)

    Wichner, R. P.; Solomon, A. D.; Drake, J. B.; Williams, P. T.

    1988-01-01

    A thermal analysis was performed of a thermal energy storage canister of a type suggested for use in a solar receiver for an orbiting Brayton cycle power system. Energy storage for the eclipse portion of the cycle is provided by the latent heat of a eutectic mixture of LiF and CaF2 contained in the canister. The chief motivation for the study is the prediction of vapor void effects on temperature profiles and the identification of possible differences between ground test data and projected behavior in microgravity. The first phase of this study is based on a two-dimensional, cylindrical coordinates model using an interim procedure for describing void behavor in 1-g and microgravity. The thermal analysis includes the effects of solidification front behavior, conduction in liquid/solid salt and canister materials, void growth and shrinkage, radiant heat transfer across the void, and convection in the melt due to Marangoni-induced flow and, in 1-g, flow due to density gradients. A number of significant differences between 1-g and o-g behavior were found. This resulted from differences in void location relative to the maximum heat flux and a significantly smaller effective conductance in 0-g due to the absence of gravity-induced convection.

  4. Generic control of material handling systems

    NARCIS (Netherlands)

    Haneyah, S.W.A.

    2013-01-01

    Material handling systems (MHSs) are in general complex installations that raise challenging design and control problems. In the literature, design and control problems have received a lot of attention within distinct business sectors or systems, but primarily from a system’s user perspective. Much

  5. Canister positioning. Influence of fracture system on deposition hole stability

    International Nuclear Information System (INIS)

    The study concerns the mechanical behaviour of rock surrounding tunnels and deposition holes in a nuclear waste repository. The mechanical effects of tunnel excavation and deposition hole excavation are investigated by use of a tunnel scale numerical model representing a part of a KBS-3 type repository. The excavation geometry, the initial pre-mining state of stress, and the geometrical features of the fracture system are defined according to conditions that prevail in the TBM tunnel rock mass in Aespoe HRL. Comparisons are made between results obtained without consideration of fractures and results obtained with inclusion of the fracture system. The focus is on the region around the intersection of a tunnel and a deposition hole. A general conclusion is that a fracture system of the type found in the TBM rock mass does not have a decisive influence on the stability of the deposition holes. To estimate the expected extent of spalling, information about other conditions, e.g. the orientation of the initial stresses and the strength properties of the intact rock, is more important than detailed information about the fracture system

  6. Engineered Barrier System - Mechanical Integrity of KBS-3 Spent Fuel Canisters. Report from a Workshop. Synthesis and extended abstracts

    International Nuclear Information System (INIS)

    SKI is preparing to review the license applications being developed by the Swedish Nuclear Fuel and Waste Management Company (SKB) for a final repository for the geological disposal of spent nuclear fuel in the year 2009. As part of its preparation, SKI is conducting a series of technical workshops on key aspects of the Engineered Barrier System (EBS). The workshop reported here mainly dealt with the mechanical integrity of KBS-3 spent fuel canisters. This included assessment and review of various loading conditions, structural integrity models and mechanical properties of the copper shell and the cast iron insert. Degradation mechanisms such as stress corrosion cracking and brittle creep fracture were also briefly addressed. Previous workshops have addressed the overall concept for long-term integrity of the EBS, the manufacturing, testing and QA of the EBS, the performance confirmation for the EBS, long-term stability of the buffer and the backfill, corrosion properties of copper canisters and the spent fuel dissolution and source term modelling. The goal of ongoing review work in connection of the workshop series is to achieve a comprehensive overview of all aspects of SKB's EBS and spent fuel work prior to the handling of the forthcoming license application. This report aims to summarise the issues discussed at the workshop and to extract the essential viewpoints that have been expressed. The report is not a comprehensive record of all the discussions at the workshop, and individual statements made by workshop participants should be regarded as personal opinions rather than SKI viewpoints. Results from the EBS workshops series will be used as one important basis in future review work. This reports includes in addition to the workshop synthesis, questions to SKB identified prior to the workshop, and extended abstracts for introductory presentations

  7. Remote controlled mover for disposal canister transfer

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

    This working report is an update for an earlier automatic guided vehicle design (Pietikaeinen 2003). The short horizontal transfers of disposal canisters manufactured in the encapsulation process are conducted with remote controlled movers both in the encapsulation plant and in the underground areas at the canister loading station of the disposal facility. The canister mover is a remote controlled transfer vehicle mobile on wheels. The handling of canisters is conducted with the assistance of transport platforms (pallets). The very small automatic guided vehicle of the earlier design was replaced with a commercial type mover. The most important reasons for this being the increased loadbearing requirement and the simpler, proven technology of the vehicle. The larger size of the vehicle induced changes to the plant layouts and in the principles for dealing with fault conditions. The selected mover is a vehicle, which is normally operated from alongside. In this application, the vehicle steering technology must be remote controlled. In addition, the area utilization must be as efficient as possible. This is why the vehicle was downsized in its outer dimensions and supplemented with certain auxiliary equipment and structures. This enables both remote controlled operation and improves the vehicle in terms of its failure tolerance. Operation of the vehicle was subjected to a risk analysis (PFMEA) and to a separate additional calculation conserning possible canister toppling risks. The total cost estimate, without value added tax for manufacturing the system amounts to 730 000 euros. (orig.)

  8. Remote controlled mover for disposal canister transfer

    International Nuclear Information System (INIS)

    This working report is an update for an earlier automatic guided vehicle design (Pietikaeinen 2003). The short horizontal transfers of disposal canisters manufactured in the encapsulation process are conducted with remote controlled movers both in the encapsulation plant and in the underground areas at the canister loading station of the disposal facility. The canister mover is a remote controlled transfer vehicle mobile on wheels. The handling of canisters is conducted with the assistance of transport platforms (pallets). The very small automatic guided vehicle of the earlier design was replaced with a commercial type mover. The most important reasons for this being the increased loadbearing requirement and the simpler, proven technology of the vehicle. The larger size of the vehicle induced changes to the plant layouts and in the principles for dealing with fault conditions. The selected mover is a vehicle, which is normally operated from alongside. In this application, the vehicle steering technology must be remote controlled. In addition, the area utilization must be as efficient as possible. This is why the vehicle was downsized in its outer dimensions and supplemented with certain auxiliary equipment and structures. This enables both remote controlled operation and improves the vehicle in terms of its failure tolerance. Operation of the vehicle was subjected to a risk analysis (PFMEA) and to a separate additional calculation conserning possible canister toppling risks. The total cost estimate, without value added tax for manufacturing the system amounts to 730 000 euros. (orig.)

  9. The remote handling systems for ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

  10. MATERIAL HANDLING IN FLEXIBLE MANUFACTURING SYSTEM

    Directory of Open Access Journals (Sweden)

    Mr. Neeraj Dahiya

    2011-08-01

    Full Text Available The objective of this study is to analyze the system performanceof a flexible manufacturing system cell. The study givesinformation on production potential of the cell by groupingcommon parts. To complete this, computer simulation models aredeveloped using the SIMAN simulation language. Initially nomaterial handling is provided to the manufacturing system to getan upper bound estimate of production output. Next, we explorethe impact that an automatic guided vehicle (AGV has on systemperformance with manufacturing system. The final analysis isperformed in which a conveyor is implemented for the materialhandling. The performance result with comparison is presented inthe form of confidence intervals. After examine the simulationresults, we recommend to implement a conveyor system formaterial handling. Use of AGV in the flexible manufacturingsystem creates a bottleneck which causes a dramatically decreasein the production: as compare to a conveyor as the materialhandling system which does not limit the daily production outputof the manufacturing cell.

  11. Filter Measurement System for Nuclear Material Storage Canisters. End of Year Report FY 2013

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reeves, Kirk P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-02-03

    A test system has been developed at Los Alamos National Laboratory to measure the aerosol collection efficiency of filters in the lids of storage canisters for special nuclear materials. Two FTS (filter test system) devices have been constructed; one will be used in the LANL TA-55 facility with lids from canisters that have stored nuclear material. The other FTS device will be used in TA-3 at the Radiation Protection Division’s Aerosol Engineering Facility. The TA-3 system will have an expanded analytical capability, compared to the TA-55 system that will be used for operational performance testing. The LANL FTS is intended to be automatic in operation, with independent instrument checks for each system component. The FTS has been described in a complete P&ID (piping and instrumentation diagram) sketch, included in this report. The TA-3 FTS system is currently in a proof-of-concept status, and TA-55 FTS is a production-quality prototype. The LANL specification for (Hagan and SAVY) storage canisters requires the filter shall “capture greater than 99.97% of 0.45-micron mean diameter dioctyl phthalate (DOP) aerosol at the rated flow with a DOP concentration of 65±15 micrograms per liter”. The percent penetration (PEN%) and pressure drop (DP) of fifteen (15) Hagan canister lids were measured by NFT Inc. (Golden, CO) over a period of time, starting in the year 2002. The Los Alamos FTS measured these quantities on June 21, 2013 and on Oct. 30, 2013. The LANL(6-21-2013) results did not statistically match the NFT Inc. data, and the LANL FTS system was re-evaluated, and the aerosol generator was replaced and the air flow measurement method was corrected. The subsequent LANL(10-30-2013) tests indicate that the PEN% results are statistically identical to the NFT Inc. results. The LANL(10-30-2013) pressure drop measurements are closer to the NFT Inc. data, but future work will be investigated. An operating procedure for the FTS (filter test system) was written, and

  12. MATERIAL HANDLING IN FLEXIBLE MANUFACTURING SYSTEM

    OpenAIRE

    Mr. Neeraj Dahiya; Mr. Neeraj Nirmal

    2011-01-01

    The objective of this study is to analyze the system performanceof a flexible manufacturing system cell. The study givesinformation on production potential of the cell by groupingcommon parts. To complete this, computer simulation models aredeveloped using the SIMAN simulation language. Initially nomaterial handling is provided to the manufacturing system to getan upper bound estimate of production output. Next, we explorethe impact that an automatic guided vehicle (AGV) has on systemperforma...

  13. DYNAMIC BOTTLENECKS IN HANDLING AND STORAGE SYSTEMS

    OpenAIRE

    PANOVA YULIA; HILMOLA OLLI-PEKKA

    2015-01-01

    The development of industrial engineering and production systems is manifested under the demand of Russian customers in the current economic and political situation, e.g. deprivation from several import markets. In these circumstances, issues related to the formation of process systems are gaining their importance. The article considers the objective of reaching the smooth and continuous material flow in the handling and storage system of the plant, as well as the problems of bottlenecks opti...

  14. Generic control of material handling systems

    OpenAIRE

    Haneyah, S.W.A.

    2013-01-01

    Material handling systems (MHSs) are in general complex installations that raise challenging design and control problems. In the literature, design and control problems have received a lot of attention within distinct business sectors or systems, but primarily from a system’s user perspective. Much less attention is paid to generic (i.e., sector independent) control architectures and modeling approaches across these various sectors, which is in the first place interesting for MHS suppliers. I...

  15. Data handling system for SST-1

    International Nuclear Information System (INIS)

    For carrying out experiments on Steady State Superconducting Tokamak-1 (SST-1) in the Institute for Plasma Research (IPR), Gandhinagar, a system for plant and experimental data handling and access is developed and has been used in the Institute since the experiments has began. The SAN based central storage system maintains the whole cycle of experimental data handling: from storing configuration data of plants and experiments systems to the acquisition of raw data from the fusion device (SST-1), to the presentation of processed data and support for the experiment and plant data archive. The storage system facilities the researchers to access the data both locally from within the experiment network and as well as remotely from various sites of the IPR campus network. The system developed is based on modern principle of SAN architecture that allows to produce and handle larger amounts of experimental data without single point of failure, thus providing the opportunities to intensify and extend the fusion researches. The features of the system along with the design principles are reviewed in this paper. (author)

  16. Airborne Effluent Monitoring System Certification for New Canister Storage Building Ventilation Exhaust Stack

    International Nuclear Information System (INIS)

    Pacific Northwest National Laboratory conducted three of the six tests needed to verify that the effluent monitoring system for the new Canister Storage Building ventilation exhaust stack meets applicable regulatory performance criteria for air sampling systems at nuclear facilities. These performance criteria address both the suitability of the location for the air-sampling probe and the transport of the sample to the collection devices. The criteria covering the location for the air-sampling probe ensure that the contaminants in the stack are well mixed with the airflow at the probe location such that the extracted sample represents the whole. The sample-transport criteria ensure that the sampled contaminants are quantitatively delivered to the collection device. The specific performance criteria are described in detail in this report. The tests reported here cover the contaminant tracer uniformity and particle delivery performance criteria. These criteria were successfully met. The other three tests were conducted by the start-up staff of Duke Engineering and Services Hanford Inc. (DESH) and reported elsewhere. The Canister Storage Building is located in the 200 East Area of the U.S. Department of Energy's Hanford Site near Richland, Washington. The new air-exhaust system was built under the W379 Project. The air sampling system features a probe with a single shrouded sampling nozzle, a sample delivery line, and a filter holder to collect the sample

  17. VVER NPPs fuel handling machine control system

    International Nuclear Information System (INIS)

    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

  18. Remote handling systems for the Pride application

    International Nuclear Information System (INIS)

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

  19. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs

  20. System Configuration Management Implementation Procedure for the Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    This document implements the procedure for providing configuration control for the monitoring and control systems associated with the operation of the Canister Storage Building (CSB). It identifies and defines the configuration items in the monitoring and control systems, provides configuration control of these items throughout the system life cycle, provides configuration status accounting, physical protection and control, and verifies the completeness and correctness of the items. It is written to comply with HNF-SD-SNF-CM-001, Spent Nuclear Fuel Configuration Management Plan (Forehand 1998), HNF-PRO-309, Computer Software Quality Assurance Requirements, HNF-PRO-2778, IRM Application Software System Life Cycle Standards, and applicable sections of administrative procedure AP-CM-6-037-00, SNF Project Process Automation Software and Equipment Configuration Management

  1. System Configuration Management Implementation Procedure for the Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    GARRISON, R.C.

    2000-09-22

    This document implements the procedure for providing configuration control for the monitoring and control systems associated with the operation of the Canister Storage Building (CSB). It identifies and defines the configuration items in the monitoring and control systems, provides configuration control of these items throughout the system life cycle, provides configuration status accounting, physical protection and control, and verifies the completeness and correctness of the items. It is written to comply with HNF-SD-SNF-CM-001, Spent Nuclear Fuel Configuration Management Plan (Forehand 1998), HNF-PRO-309, Computer Software Quality Assurance Requirements, HNF-PRO-2778, IRM Application Software System Life Cycle Standards, and applicable sections of administrative procedure AP-CM-6-037-00, SNF Project Process Automation Software and Equipment Configuration Management.

  2. Acceptance of canisters of high-level waste by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high-level waste will be accepted. This document discusses the acceptance of canisters of high-level waste by the Federal Waste Management System. 16 refs., 7 figs., 11 tabs

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. WWER NPPs fuel handling machine control system

    International Nuclear Information System (INIS)

    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

  6. Canister cryogenic system for cooling germanium semiconductor detectors in borehole and marine probes

    Science.gov (United States)

    Boynton, G.R.

    1975-01-01

    High resolution intrinsic and lithium-drifted germanium gamma-ray detectors operate at about 77-90 K. A cryostat for borehole and marine applications has been designed that makes use of prefrozen propane canisters. Uses of such canisters simplifies cryostat construction, and the rapid exchange of canisters greatly reduces the time required to restore the detector to full holding-time capability and enhances the safety of a field operation where high-intensity 252Cf or other isotopic sources are used. A holding time of 6 h at 86 K was achieved in the laboratory in a simulated borehole probe in which a canister 3.7 cm diameter by 57 cm long was used. Longer holding times can be achieved by larger volume canisters in marine probes. ?? 1975.

  7. 2D and 3D thermal simulations for storage systems with internal natural convection for canistered spent fuel

    International Nuclear Information System (INIS)

    In the US, the number of nuclear plants expected to implement on-site dry storage is increasing each year. As reactors burn advanced fuel assemblies to higher burnups, the dry storage systems will be required to accommodate higher heat loads. This is due to the increasing capacity of the systems and the need to store higher burnup fuel with reasonable cooling periods (i.e., five to six years). As the storage systems heat rejection design must be passive, natural convection is an efficient means for rejection of heat from the spent fuel to the surface of the canister boundary. The design presented in this paper is a canistered system that employs conduction, radiation and convection to reject heat from the canister, which is stored in a vertical concrete cask. The canister containing the spent fuel in this design is a right circular stainless steel vessel capable of storing 37 PWR fuel assemblies with a total canister heat load of 40 kW. Accompanying any design effort is the use of a numerical methodology that can accurately predict the peak-clad temperatures of the fuel and the structural components of the system. The main challenge to any analysis employing internal natural convection may be perceived as a practical limitation due to the size of the model. Since canisters are typically cylindrical, a two-dimensional model can be used to represent the canister. The fuel basket structure, which maintains the configuration of the spent fuel, is an array of square tubes, and is non-axisymmetric. Flow up through the fuel region in the basket encounters a complex cross section due to the fuel assembly rod array (up to 17 x 17). The flow region of the heated gas down the outside of the basket in the annulus between the canister shell and the basket assembly (downcomer) is also an irregular shaped area. To confirm that a two-dimensional (2D) modelling methodology is appropriate, a benchmark using results from a thermal test is required. The thermal test focuses on the

  8. Development of an alternative plutonium canister assay system (APCA) using He-3 alternative neutron detector

    International Nuclear Information System (INIS)

    In order to deal with the global shortage of He-3 gas, He-3 alternative neutron detectors using ZnS/10B2O3 ceramic scintillators for nuclear security and the safeguards, and a demonstrator of Alternative Plutonium Canister Assay System (APCA) for the safeguards NDA in which the alternative detectors are employed, have been developed with the support of Japanese government (the Ministry of Education, Culture, Sports, Science and Technology). The results of the optical guide property of scintillation lights in the alternative detector tubes derived from the simulations using a ray-tracing code are presented in comparison with the test results of the developed alternative detectors. Furthermore, the fundamental performance of APCA estimated from the neutron Monte-Carlo code MVP and the comparison with the performance of the current PCAS are also described, respectively, together with the future plan of the APCA demonstration test. (author)

  9. Retrievability of spent nuclear fuel canisters; Kaeytetyn ydinpolttoaineen loppusijoituskapseleiden palautettavuus

    Energy Technology Data Exchange (ETDEWEB)

    Saanio, T. [Saanio and Riekkola Oy, Helsinki (Finland); Raiko, H. [VTT Energy, Espoo (Finland)

    1999-03-01

    As a part of the designing process of the Finnish spent nuclear fuel repository, a preliminary study has been carried out to investigate how the canisters could technically be retrieved to the ground surface. Possibility of retrieving a canister has been investigated in different phases of the disposal project. Retrievability has not been a design goal for the spent fuel repository. However, design of the repository includes some features that may ease the retrieval of canisters in the future. Spent fuel elements are packaged in massive copper-iron canisters, which are mechanically strong and long-lived. The repository consists of excavated tunnels in hard rock which are supposed to be very long-lived making the removal of the tunnel backfilling technically possible also in the future. As long as the bentonite buffer has not been installed the canister can be returned to the ground surface using the same equipment as was used when the canister was brought down to the repository and lowered into the hole. In the encapsulation station the spent fuel elements can be packaged in the other canister or in the transport cask. After a deposition tunnel has been backfilled and closed, the retrieval consists of tearing down the concrete structure at the entry of the deposition tunnel, removal of the tunnel backfilling, removal of the bentonite from the disposal hole and lifting up of the canister. Various methods, e.g., flushing the bentonite with saline solutions, can be used to detach the canister from a hole with fully saturated bentonite. Recovery will be technically possible also after closing of the disposal facility. Backfilling of the shafts and tunnels will be removed and additional new structures and systems will have to be built in the repository. After that canisters can be transported to the ground surface as described above. In addition, handling of the canisters at the ground surface will require additional facilities. Canisters can be packaged in the

  10. Handling Occlusions for Robust Augmented Reality Systems

    Directory of Open Access Journals (Sweden)

    Maidi Madjid

    2010-01-01

    Full Text Available Abstract In Augmented Reality applications, the human perception is enhanced with computer-generated graphics. These graphics must be exactly registered to real objects in the scene and this requires an effective Augmented Reality system to track the user's viewpoint. In this paper, a robust tracking algorithm based on coded fiducials is presented. Square targets are identified and pose parameters are computed using a hybrid approach based on a direct method combined with the Kalman filter. An important factor for providing a robust Augmented Reality system is the correct handling of targets occlusions by real scene elements. To overcome tracking failure due to occlusions, we extend our method using an optical flow approach to track visible points and maintain virtual graphics overlaying when targets are not identified. Our proposed real-time algorithm is tested with different camera viewpoints under various image conditions and shows to be accurate and robust.

  11. CERN Sells its Electronic Document Handling System

    CERN Multimedia

    2001-01-01

    The EDH team. Left to right: Derek Mathieson, Rotislav Titov, Per Gunnar Jonsson, Ivica Dobrovicova, James Purvis. Missing from the photo is Jurgen De Jonghe. In a 1 MCHF deal announced this week, the British company Transacsys bought the rights to CERN's Electronic Document Handling (EDH) system, which has revolutionised the Laboratory's administrative procedures over the last decade. Under the deal, CERN and Transacsys will collaborate on developing EDH over the coming 12 months. CERN will provide manpower and expertise and will retain the rights to use EDH, which will also be available freely to other particle physics laboratories. This development is an excellent example of the active technology transfer policy CERN is currently pursuing. The negotiations were carried out through a fruitful collaboration between AS and ETT Divisions, following the recommendations of the Technology Advisory Board, and with the help of SPL Division. EDH was born in 1991 when John Ferguson and Achille Petrilli of AS Divisi...

  12. Copper canister with cast inner component. Amendment to project on Alternative Systems Study (PASS), SKB TR 93-04

    International Nuclear Information System (INIS)

    The Project on Alternative Systems Study, PASS, was described in a report dated october 1992. In the report, the reference repository concept KBS-3 is described together with three other alternatives. In the report several designs for fuel storage canisters are presented. This report describes a recently developed design for the inner component of the composite, steel and copper, canister which is the main alternative in the KBS-3-model. The new design will be manufactured by casting. A cast insert with inner walls eliminates the need for a stabilizing filler in the canister and guarantees that the fuel remains sub-critical during sufficient time in the repository. The cast insert is judged, to, in comparison with the steel tube alternative, lead to a considerably simplified process in the encapsulation plant and lower development and investment cost. Positive side effects of the design are that the mechanical strength is improved by a factor 2-3 and that the difficult filling operation is avoided in the encapsulation process. The drawbacks are higher weight and probably higher unit price for the empty canister

  13. The NOAO KOSMOS Data Handling System

    CERN Document Server

    Seaman, Rob

    2015-01-01

    KOSMOS and COSMOS are twin high-efficiency imaging spectrographs that have been deployed as NOAO facility instruments for the Mayall 4-meter telescope on Kitt Peak in Arizona and for the Blanco telescope on Cerro Tololo in Chile, respectively. The NOAO Data Handling System (DHS) has seen aggressive use over several years at both the Blanco and Mayall telescopes with NEWFIRM (the NOAO Extremely Wide-Field Infrared Imager) and the Mosaic-1.1 wide-field optical imager. Both of these instruments also rely on the Monsoon array controller and related software, and on instrument-specific versions of the NOAO Observation Control System (NOCS). NOCS, Monsoon and DHS are thus a well-tested software suite that was adopted by the KOSMOS project. This document describes the specifics of the KOSMOS implementation of DHS, in particular in support of the original two-amplifier e2v 2Kx4K CCD detectors with which the instruments were commissioned. The emphasis will be on the general layout of the DHS software components and th...

  14. Aerobot Sampling and Handling System Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Honeybee Robotics proposes to: ?Derive and document the functional and technical requirements for Aerobot surface sampling and sample handling across a range of...

  15. Canister transfer in access tunnel. Lay-out, system and operation description

    International Nuclear Information System (INIS)

    In this report the alternative of canister transfer by a vehicle is examined, the principle and the plans are shown in those details that differ from the canister-transfer-throughshaft alternative. In vehicle-transfer alternative the disposal canisters are transferred with a freely steered motor vehicle from ground surface to the repository at level 400 to 500 m below ground surface. The vehicle is a crawler type heavy-load transfer vehicle. The disposal canisters are loaded into the shield cylinder of the vehicle at the encapsulation plant. Canisters are transferred with the vehicle from encapsulation plant to the mouth of the repository ramp, then through the ramp to the repository level underground and finally through central tunnels to the disposal tunnel and disposal hole. Radiation effects of the canister can be detected only in the close vicinity of the vehicle. Transfer route in the site area shall be selected in a way that heavy traffic areas are avoided and the roads used should be even and passable. Underground, the canister transfer proceeds always in the controlled area. The access ramp is declared to be controlled area temporarily in four sections as the transfer proceeds through the ramp. The ventilation is temporarily closed in the controlled area section during canister transfer. To transfer the vehicle from access ramp to the technical rooms of the controlled area of the repository level a construction of a by-pass tunnel is planned. This is made for avoiding disturbance of the simultaneous uncontrolled area operations on the repository level. In two-storey alternative, a by-pass tunnel access is needed also on the lower level of the repository. In case of one-storey repository alternative, the vehicle transfer of the disposal canister does not cause any changes in the order of use of the disposal tunnels or in the organization of controlled and uncontrolled area. In case of two-storey repository, the order of the use of some disposal tunnels is

  16. A Knowledge-Based Approach for Selection of Material Handling Equipment and Material Handling System Pre-design

    OpenAIRE

    YAMAN, Ramazan

    2001-01-01

    For material handling system design, material handling equipment selection is the first stage. Also the material handling system and facility layout design problems are coupled. Solving these problems needs consideration of these three different problems. Right material handling equipment selection and good design of the material handling system and facility layout can increase productivity and reduce investments and operations' costs. In this study, after describing the mater...

  17. Application Examples for Handle System Usage

    Science.gov (United States)

    Toussaint, F.; Weigel, T.; Thiemann, H.; Höck, H.; Stockhause, M.; Lautenschlager, M.

    2012-12-01

    Besides the well-known DOI (Digital Object Identifiers) as a special form of Handles that resolve to scientific publications there are various other applications in use. Others perhaps are just not yet. We present some examples for the existing ones and some ideas for the future. The national German project C3-Grid provides a framework to implement a first solution for provenance tracing and explore unforeseen implications. Though project-specific, the high-level architecture is generic and represents well a common notion of data derivation. Users select one or many input datasets and a workflow software module (an agent in this context) to execute on the data. The output data is deposited in a repository to be delivered to the user. All data is accompanied by an XML metadata document. All input and output data, metadata and the workflow module receive Handles and are linked together to establish a directed acyclic graph of derived data objects and involved agents. Data that has been modified by a workflow module is linked to its predecessor data and the workflow module involved. Version control systems such as svn or git provide Internet access to software repositories using URLs. To refer to a specific state of the source code of for instance a C3 workflow module, it is sufficient to reference the URL to the svn revision or git hash. In consequence, individual revisions and the repository as a whole receive PIDs. Moreover, the revision specific PIDs are linked to their respective predecessors and become part of the provenance graph. Another example for usage of PIDs in a current major project is given in EUDAT (European Data Infrastructure) which will link scientific data of several research communities together. In many fields it is necessary to provide data objects at multiple locations for a variety of applications. To ensure consistency, not only the master of a data object but also its copies shall be provided with a PID. To verify transaction safety and to

  18. High-level waste canister storage final design, installation, and testing. Topical report

    International Nuclear Information System (INIS)

    This report is a description of the West Valley Demonstration Project's radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project's vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access

  19. High-level waste canister storage final design, installation, and testing. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

    1998-04-01

    This report is a description of the West Valley Demonstration Project`s radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project`s vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access.

  20. Experience in handling abnormal and emergency situations in PHWR fuel handling system

    International Nuclear Information System (INIS)

    On-power Fuel Handling System of PHWR reactor consists of complicated mechanisms operating in multiple media like heavy water, light water and oil. This remote controlled system is the lifeline of PHWR reactor. The complexity of on-power fuel handling system and the need to continuously improve its performance presents challenges at every step. A large number of innovations, modifications and improvements in the system have been made by the stations, design group and R and D units to meet the challenges of higher refueling rate. Innovations in operating/maintenance practices and the methods to safely retrieve from abnormal/emergency situations in shortest possible time had to be specifically devised from the embryonic stage. A lot of efforts were required to be put in by various agencies to develop and formalise the operating procedures for handling various emergency conditions. The implementation of these procedures required the development of special tools/fixtures which had to be tested and tried out in mock-ups before their actual use. The retrieval from emergency situations like handling of damaged bundles in MAPS in early eighties, bundles dropped in shuttle station in NAPS in 1998 and failure of fuel string to move due to damaged bundles at Kaiga in 2003 are some of the most difficult situations handled over the years.This paper focuses on the challenges faced during handling of Safety-related Events in PHWR Fuel Handling System. It also discusses development of procedures and tooling to retrieve from abnormal situations and various innovations and design improvements to avoid the recurrence of the events. (author)

  1. Evaluation of radiation shielding performance of disposal canister storing PWR spent fuels

    International Nuclear Information System (INIS)

    Radiation shielding is an important factor in designing disposal canister containing spent nuclear fuel(SNF), because intensity for photon and neutron in SNF assembly after 40 year cooling is still high, ∼1015 photons/sec and ∼108 neutrons/sec, respectively. Radiation escaping from the disposal canister emplaced in repository may cause radiolysis and form oxidizing chemical species. This may result in corrosion of canister itself to proceed. Personnel exposure is also important concern. If shielding performance of canister can reduce radiation level to 1mRem/hr, human access without a control on duration and frequency of exposure may be possible. This provides the benefit of more direct human control of waste packages handling and emplacement operations. In this paper, the radiation shielding performance was evaluated based on current reference disposal system to check absorbed dose for radiolysis, and exposure dose for radiation protection

  2. Equipment for deployment of canisters with spent nuclear fuel and bentonite buffer in horizontal holes

    International Nuclear Information System (INIS)

    The study presents the predesign of equipment for the deployment of canisters in long horizontal holes. The canisters are placed in the centre of the hole and are surrounded by a bentonite buffer. In thE study the canisters are assumed to have a diameter of 1.6 m and a length of 5.9 m, including the hemispherical ends. Their total weight is 60 tonnes. The bentonite buffer after homogenization is 400 mm thick, making a total package diameter of 2.4 m. The deployment system consists of four wagons for handling The canisters and the bentonite blocks. To ensure safe emplacement, every part is installed separately in its final position. This also makes it possible to use small clearances between the canisters and the bentonite blocks and between the blocks and the rock wall. With small clearances, backfilling can be avoided. Another basic design idea is that the wagons are equipped with wheels, which are in direct contact with the rock walls. Thus, rails, which have to be removed as the deployment progresses, are unnecessary. To minimize the time taken for deploying one canister, the wagons are designed so that only three trips from the service area to the deposit area are needed. Due to the radiation in the vicinity of the canisters, the wagons have to be teleoperated

  3. Equipment for deployment of canisters with spent nuclear fuel and bentonite buffer in horisontal holes

    International Nuclear Information System (INIS)

    This study presents the predesign of equipment for the deployment of canisters in long horizontal holes. The canisters are placed in the centre of the hole and are surrounded by a bentonite buffer. In this study the canisters are assumed to have a diameter of 1.6 m and a length of 5.9 m, including the hemispherical ends. Their total weight is 60 tonnes. The bentonite buffer after homogenization is 400 mm thick, making a total package diameter of 2.4 m. The deployment system consists of four wagons for handling the canisters and the bentonite blocks. To ensure safe emplacement, every part is installed separately in its final position. This also makes it possible to use small clearances between the canisters and the bentonite blocks and between the blocks and the rock wall. With small clearances, backfilling can be avoided. Another basic design idea is that the wagons are equipped with wheels, which are in direct contact with the rock walls. Thus, rails, which have to be removed as the deployment progresses, are unnecessary. To minimize the time taken for deploying one canister, the wagons are designed so that only three trips from the service area to the deposit area are needed. Due to the radiation in the vicinity of the canisters, the wagons have to be teleoperated. (au)

  4. Mechanical handling systems in the Sellefield vitrification plant

    International Nuclear Information System (INIS)

    British Nuclear Fuels plc (BNFL) has over 40 years experience in the design, construction and operation of nuclear reprocessing plants and of waste management. Many of these plants have required extensive mechanical handling systems, the handling and control systems designed and developed for the Sellafield Vitrification Plant and Product Store are described. These systems are now fully operational and illustrate many of the features and techniques developed by BNFL for nuclear package handling. Utilisation of these systems and similar systems in other Sellafield Plants has demonstrated notable advantages in ease/flexibility of operations, product quality and costs. (author)

  5. 324 Building Liquid Waste Handling System Functional Design Criteria

    International Nuclear Information System (INIS)

    The 324 Building in the 300 Area of the Hanford Site, is preparing to design, construct, and operate the Liquid Waste Handling System (LWHS). The system will include transfer, collection, treatment, and disposal of radiological and mixed liquid waste

  6. CHLOE: a system for the automatic handling of spark pictures

    International Nuclear Information System (INIS)

    The system for automatic data handling uses commercially available or state-of-the-art components. The system is flexible enough to accept information from various types of experiments involving photographic data acquisition

  7. Handling effluent from nuclear thermal propulsion system ground tests

    International Nuclear Information System (INIS)

    A variety of approaches for handling effluent from nuclear thermal propulsion system ground tests in an environmentally acceptable manner are discussed. The functional requirements of effluent treatment are defined and concept options are presented within the framework of these requirements. System concepts differ primarily in the choice of fission-product retention and waste handling concepts. The concept options considered range from closed cycle (venting the exhaust to a closed volume or recirculating the hydrogen in a closed loop) to open cycle (real time processing and venting of the effluent). This paper reviews the different methods to handle effluent from nuclear thermal propulsion system ground tests

  8. Conceptual designs of radioactive canister transporters

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-02-01

    This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts.

  9. Conceptual designs of radioactive canister transporters

    International Nuclear Information System (INIS)

    This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts

  10. Advanced handling-systems with enhanced performance flexibility

    International Nuclear Information System (INIS)

    This report describes the results of a project related to future applications and requirements for advanced handling systems. This report consists of six chapters. Following the description of the aims the tools for setting up the requirements for the handling systems including the experience during the data acquisition process is described. Furthermore some information is given about the current state of the art of robotics and manipulators. Of paramount importance are the descriptions of applications and related concepts in the following chapters leading to specific categories of advanced handling units. The paper closes with the description of the first concepts for realization. (orig./HP)

  11. 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.

  12. Monitored Retrievable Storage/Multi-Purpose Canister analysis: Simulation and economics of automation

    International Nuclear Information System (INIS)

    Robotic automation is examined as a possible alternative to manual spent nuclear fuel, transport cask and Multi-Purpose canister (MPC) handling at a Monitored Retrievable Storage (MRS) facility. Automation of key operational aspects for the MRS/MPC system are analyzed to determine equipment requirements, through-put times and equipment costs is described. The economic and radiation dose impacts resulting from this automation are compared to manual handling methods

  13. SLSF loop handling system. Volume I. Structural analysis

    International Nuclear Information System (INIS)

    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

  14. A graphics based remote handling control system

    International Nuclear Information System (INIS)

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

  15. Enbridge system : crude types, transportation and handling systems

    Energy Technology Data Exchange (ETDEWEB)

    Anand, A. [Enbridge Corp., Edmonton, AB (Canada)

    2009-07-01

    The supply of crude oil from the Western Canada Sedimentary Basin is expected to increase by approximately 2.1 million barrels per day by 2015. The crudes that Enbridge handles range from 19 API to 40 API and 0.1 per cent sulphur to 4.7 per cent sulphur. The diverse supply of crude oil that the Enbridge system handles includes conventional heavy, synthetic heavy, heavy high tan, heavy low residual, medium, light sour, heavy sour, light sweet, light sweet synthetic, condensate and olefinic crudes. This presentation discussed Enbridge's plans for infrastructure expansion, crude types and quality assurance program. The company's infrastructure plans include the expansion of regional pipelines to bring more supplies to the mainline; expansion of the mainline capacity to existing markets; and providing pipeline access to new markets. Merchant storage terminals will be provided in some locations. The quality of various crude types will be maintained through judicious sequencing and tank bottoms crossings. tabs., figs.

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

    International Nuclear Information System (INIS)

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

  17. Evolution of the Darlington NGS fuel handling computer systems

    International Nuclear Information System (INIS)

    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

  18. Sensor Based Effective Monitoring of Coal Handling System (CHS

    Directory of Open Access Journals (Sweden)

    Kuttalakkani.M

    2013-06-01

    Full Text Available Coal level detection is an important aspect to assess the performance of a coal-fired power plant. Coal has to be transported, via a coal handling system. The fuel in a coal-fired power plant is stored in silos, bunkers or stock piles. Coal is stored in silos in a small plant, Bunkers for handling a day’s operation and Stock piling methods for large plants. So, fuel handling had to done efficiently. To accurately sense the coal height, Real-time feedback is deployed within the bunker or stock pile. The real time range information is then fedback to the control system. Of the different types of ranging sensors, radar based system is used. Also a real-time temperature monitoring system is developed to protect the coal. The range and temperature data from sensors are sent to the main system through GSM modem by means of SMS. The range information is used to start the conveyor belt to draw the coal from coal yard. If the temperature exceeds the limit, the SMS will be sent through the software or it will call the respective person to monitor the process. A fire sensor is also used to extinguish the fire by initiating the water spraying system. A PIC Microcontroller is interfaced all the sensors for effective handling of thermal power plant.

  19. Process and machinery description of equipment for deposition of canisters in horizontal deposition holes

    International Nuclear Information System (INIS)

    In this report are presented seventeen methods to deposit canisters with spent nuclear fuel in horizontal holes, one canister per hole, in the KBS-3 system. They have been developed successively, after an analysis of weak points and strong points in previously described methods. In conformance with the guidelines for Project JADE, two choices of system have been considered during the development work. One choice is whether the canister should be provided with a tubular radiation shield or not during transport in the secondary tunnels. Another choice is whether canister and bentonite buffer should be deposited at different occasions, but shortly after each other ('in parts') or together in a single package ('in a package'). The basic technical problem is placing heavy objects, the canister and the buffer components, in an horizontal hole which is 8 m long. For depositing of bentonite buffer and canister 'in parts', the use of a guiding pipe has been studied to reduce the impact of a sliding canister on the bentonite rings. For depositing 'in a package', three alternative techniques have been studied: a loading laddle that is rotated, a fork carriage and rails. Development has been aimed at avoiding the use of a guiding pipe and at reducing the cross section area of the secondary tunnel. A failure mode and effect analysis has been performed for three of the methods in order to provide a basis for a decision whether to use a tubular radiation shield around the canister during transport and handling in the secondary tunnels. SKB has subsequently decided, partly on this basis, that the canisters should be placed in radiation shields. The development work reported here has not yet yielded a definitive method for placing canisters in horizontal holes. It is recommended that in the continued work: canister and bentonite buffer are deposited in a hole at the same time, as a package; methods involving a minimum number of movements in the tunnel are preferred and that

  20. Interdisciplinary Design of a Pervasive Fall Handling System

    OpenAIRE

    VAN DEN BERGH, Jan; Luyten, Kris; Elprama, Shirley; Jacobs, An; Aendekerk, Brenda; De Backere, Femke

    2014-01-01

    Falls among elderly are an important concern as they impact the capability to live independently. Falls do not only have a negative impact on one’s physical well-being, an increased risk of falling also has an important impact on one's psychological well-being. A context-aware fall handling system can mitigate many of the problems of falls by facilitating timely and appropriate handling of falls. In this paper, we present the results of an early exploration of using context as part of fal...

  1. Design package for fuel retrieval system fuel handling tool modification

    International Nuclear Information System (INIS)

    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

  2. Advanced robotic remote handling system for reactor dismantlement

    International Nuclear Information System (INIS)

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

  3. Decentralized Control Performances of an Experimental Web Handling System

    OpenAIRE

    Nicola Ivan Giannoccaro; Takeshi Nishida; Tetsuzo Sakamoto

    2012-01-01

    Robust and good tracking control of the speed and the tension in web handling systems in spite of changes of set point is surely one of the important challenges in the web transport systems future development. In this paper, the authors experimentally demonstrate the real applicability of a decentralized robust control to a multi‐span web transport system, which is composed of twelve guide rollers, four main sections mutually interconnected with each other. The overlapping methodology has bee...

  4. Protecting worker health and safety using remote handling systems

    International Nuclear Information System (INIS)

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

  5. Feasibility study of CANDU-9 fuel handling system

    International Nuclear Information System (INIS)

    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

  6. Application of Handle System in Chinese DOI System%Handle System在中文DOI系统中的应用

    Institute of Scientific and Technical Information of China (English)

    徐健

    2008-01-01

    概述数字对象唯一标识符(DOI)的发展历史及应用现状,介绍DOI系统的核心Handle System,包括Handie System的名称空间、数据模型以及体系结构等.在此基础上,分析Handle System在中文DOI系统中的应用,解释中文DOI系统与全球handle注册中心(GHR)、本地handle服务(LHS)之间的关系,描述中文DOI系统设计和功能,重点研究基于Handle System的DOI解析机制;最后指出中文DOI系统目前存在的问题.

  7. JOYO operation support system 'JOYCAT' based on intelligent alarm handling

    International Nuclear Information System (INIS)

    An operation support system for the experimental fast reactor 'JOYO' was developed based on an intelligent alarm-handling. A specific feature of this system, called JOYCAT (JOYO Consulting and Analyzing Tool), is in its sequential processing structure that a uniform treatment by using design knowledge base is firstly applied for all activated alarms, and an exceptional treatment by using heuristic knowledge base is then applied only for the former results. This enables us to achieve real-time and flexible alarm-handling. The first alarm-handling determines the candidates of causal alarms, important alarms with which the operator should firstly cope, through identifying the cause-consequence relations among alarms based on the design knowledge base in which importance and activating conditions are described for each of 640 alarms in a frame format. The second alarm-handling makes the final judgement with the candidates by using the heuristic knowledge base described as production rules. Then, operation manuals concerning the most important alarms are displayed to operators. JOYCAT has been in commission since September of 1990, after a wide scope of validation tests by using an on-site full-scope training simulator. (author)

  8. Advanced Information Processing System - Fault detection and error handling

    Science.gov (United States)

    Lala, J. H.

    1985-01-01

    The Advanced Information Processing System (AIPS) is designed to provide a fault tolerant and damage tolerant data processing architecture for a broad range of aerospace vehicles, including tactical and transport aircraft, and manned and autonomous spacecraft. A proof-of-concept (POC) system is now in the detailed design and fabrication phase. This paper gives an overview of a preliminary fault detection and error handling philosophy in AIPS.

  9. CLASSIFICATION OF THE MGR WASTE HANDLING BUILDING SYSTEM

    International Nuclear Information System (INIS)

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) waste handling building system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333PY ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  10. CLASSIFICATION OF THE MGR WASTE HANDLING BUILDING FIRE PROTECTION SYSTEM

    International Nuclear Information System (INIS)

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) waste handling building fire protection system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333PY ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  11. CLASSIFICATION OF THE MGR WASTE HANDLING BUILDING ELECTRICAL SYSTEM

    International Nuclear Information System (INIS)

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) waste handling building electrical system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  12. CLASSIFICATION OF THE MGR SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM

    International Nuclear Information System (INIS)

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) site-generated radiological waste handling system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  13. Contamination confinement system of irradiated materials handling laboratories

    International Nuclear Information System (INIS)

    A study to prevent radioctivity release in lab scale is presented. As a basis for the design all the limits established by the IAEA for ventilation systems were observed. An evaluation of the different parameters involved in the design have been made, resulting in the especification of the working areas, ducts and filtering systems in order to get the best conditions for the safe handling of irradiated materials. (author)

  14. Robotic control architecture development for automated nuclear material handling systems

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory (LLNL) is engaged in developing automated systems for handling materials for mixed waste treatment, nuclear pyrochemical processing, and weapon components disassembly. In support of these application areas there is an extensive robotic development program. This paper will describe the portion of this effort at LLNL devoted to control system architecture development, and review two applications currently being implemented which incorporate these technologies

  15. Analysis of K west basin canister gas

    Energy Technology Data Exchange (ETDEWEB)

    Trimble, D.J., Fluor Daniel Hanford

    1997-03-06

    Gas and Liquid samples have been collected from a selection of the approximately 3,820 spent fuel storage canisters in the K West Basin. The samples were taken to characterize the contents of the gas and water in the canisters providing source term information for two subprojects of the Spent Nuclear Fuel Project (SNFP) (Fulton 1994): the K Basins Integrated Water Treatment System Subproject (Ball 1996) and the K Basins Fuel Retrieval System Subproject (Waymire 1996). The barrels of ten canisters were sampled for gas and liquid in 1995, and 50 canisters were sampled in a second campaign in 1996. The analysis results from the first campaign have been reported (Trimble 1995a, 1995b, 1996a, 1996b). The analysis results from the second campaign liquid samples have been documented (Trimble and Welsh 1997; Trimble 1997). This report documents the results for the gas samples from the second campaign and evaluates all gas data in terms of expected releases when opening the canisters for SNFP activities. The fuel storage canisters consist of two closed and sealed barrels, each with a gas trap. The barrels are attached at a trunion to make a canister, but are otherwise independent (Figure 1). Each barrel contains up to seven N Reactor fuel element assemblies. A gas space of nitrogen was established in the top 2.2 to 2.5 inches (5.6 to 6.4 cm) of each barrel. Many of the fuel elements were damaged allowing the metallic uranium fuel to be corroded by the canister water. The corrosion releases fission products and generates hydrogen gas. The released gas mixes with the gas-space gas and excess gas passes through the gas trap into the basin water. The canister design does not allow canister water to be exchanged with basin water.

  16. A Smartphone Controlled Handheld Microfluidic Liquid Handling System

    CERN Document Server

    Li, Baichen; Guan, Allan; Dong, Quan; Ruan, Kangcheng; Hu, Ronggui; Li, Zhenyu

    2014-01-01

    Microfluidics and lab-on-a-chip technologies have made it possible to manipulate small volume liquids with unprecedented resolution, automation and integration. However, most current microfluidic systems still rely on bulky off-chip infrastructures such as compressed pressure sources, syringe pumps and computers to achieve complex liquid manipulation functions. Here, we present a handheld automated microfluidic liquid handling system controlled by a smartphone, which is enabled by combining elastomeric on-chip valves and a compact pneumatic system. As a demonstration, we show that the system can automatically perform all the liquid handling steps of a bead-based sandwich immunoassay on a multi-layer PDMS chip without any human intervention. The footprint of the system is 6 by 10.5 by 16.5cm, and the total weight is 829g including battery. Powered by a 12.8V 1500mAh Li battery, the system consumed 2.2W on average during the immunoassay and lasted for 8.7 hrs. This handheld microfluidic liquid handling platform...

  17. Design of systems for handling radioactive ion exchange resin beads

    International Nuclear Information System (INIS)

    The flow of slurries in pipes is a complex phenomenon. There are little slurry data available on which to base the design of systems for radioactive ion exchange resin beads and, as a result, the designs vary markedly in operating plants. With several plants on-line, the opportunity now exists to evaluate the designs of systems handling high activity spent resin beads. Results of testing at Robbins and Meyers Pump Division to quantify the behavior of resin bead slurries are presented. These tests evaluated the following slurry parameters; resin slurry velocity, pressure drop, bead degradation, and slurry concentration effects. A discussion of the general characteristics of resin bead slurries is presented along with a correlation to enable the designer to establish the proper flowrate for a given slurry composition and flow regime as a function of line size. Guidelines to follow in designing a resin handling system are presented

  18. Structural Sensitivity of Dry Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Karri, Naveen K.; Adkins, Harold E.; Hanson, Brady D.

    2013-09-27

    This LS-DYNA modeling study evaluated a generic used nuclear fuel vertical dry storage cask system under tip-over, handling drop, and seismic load cases to determine the sensitivity of the canister containment boundary to these loads. The goal was to quantify the expected failure margins to gain insight into what material changes over the extended long-term storage lifetime could have the most influence on the security of the containment boundary. It was determined that the tip-over case offers a strong challenge to the containment boundary, and identifies one significant material knowledge gap, the behavior of welded stainless steel joints under high-strain-rate conditions. High strain rates are expected to increase the material’s effective yield strength and ultimate strength, and may decrease its ductility. Determining and accounting for this behavior could potentially reverse the model prediction of a containment boundary failure at the canister lid weld. It must be emphasized that this predicted containment failure is an artifact of the generic system modeled. Vendor specific designs analyze for cask tip-over and these analyses are reviewed and approved by the Nuclear Regulatory Commission. Another location of sensitivity of the containment boundary is the weld between the base plate and the canister shell. Peak stresses at this location predict plastic strains through the whole thickness of the welded material. This makes the base plate weld an important location for material study. This location is also susceptible to high strain rates, and accurately accounting for the material behavior under these conditions could have a significant effect on the predicted performance of the containment boundary. The handling drop case was largely benign to the containment boundary, with just localized plastic strains predicted on the outer surfaces of wall sections. It would take unusual changes in the handling drop scenario to harm the containment boundary, such as

  19. Age and condition assessments for fuel handling systems

    International Nuclear Information System (INIS)

    Throughout the nuclear industry, power plants are approaching, or in some cases have already exceeded, their originally forecasted operation life expectancy. This situation has forced nuclear power plant operators to shift greater amounts of their focus toward equipment reliability and maintenance, while simultaneously supporting increased power generation requirements, and reductions in forced outages. As a result, support organizations for the power plants, such as maintenance, spare parts, procurement, and engineering, are under more pressure than ever. In an effort to deal with this situation, programs are being developed to help manage the effects on critical components of age-related degradation. During recent years, GE Hitachi Nuclear Energy Canada(GEH-C) has been commissioned by both Bruce Power and Ontario Power Generation to perform assessments of their Fuel Handling systems. The Fuel Handling systems were identified as fundamentally different from other plant systems, and GEH-C, as the Fuel Handling OEM, was best positioned to perform these assessments. Due to the size of the plants, and the quantity of components in the system, the scopes of the assessments covered portions of the Life Cycle Management Programs which were different for each plant. The methodologies used varied, and evolved with lessons learned as the assessments progressed. In common was a desire by the plants to categorize the levels of risk posed by different components, and to identify their criticality to the plant in terms of both safety and operating performance. The challenge for the operating stations is to take information from these assessments and incorporate it into dynamic, living programs that can be used effectively to extend operational life. The presentation outlines the key aspects of age and condition-based assessments on Fuel Handling systems, and the lessons learned that are critical to the success of these initiatives.

  20. SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    S. C. Khamankar

    2000-06-20

    The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

  1. SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

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

    International Nuclear Information System (INIS)

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

  3. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

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

  4. Adaptive and energy efficient SMA-based handling systems

    Science.gov (United States)

    Motzki, P.; Kunze, J.; Holz, B.; York, A.; Seelecke, S.

    2015-04-01

    Shape Memory Alloys (SMA's) are known as actuators with very high energy density. This fact allows for the construction of very light weight and energy-efficient systems. In the field of material handling and automated assembly process, the avoidance of big moments of inertia in robots and kinematic units is essential. High inertial forces require bigger and stronger robot actuators and thus higher energy consumption and costs. For material handling in assembly processes, many different individual grippers for various work piece geometries are used. If one robot has to handle different work pieces, the gripper has to be exchanged and the assembly process is interrupted, which results in higher costs. In this paper, the advantages of using high energy density Shape Memory Alloy actuators in applications of material-handling and gripping-technology are explored. In particular, light-weight SMA actuated prototypes of an adaptive end-effector and a vacuum-gripper are constructed via rapid-prototyping and evaluated. The adaptive end-effector can change its configuration according to the work piece geometry and allows the handling of multiple different shaped objects without exchanging gripper tooling. SMA wires are used to move four independent arms, each arm adds one degree of freedom to the kinematic unit. At the tips of these end-effector arms, SMA-activated suction cups can be installed. The suction cup prototypes are developed separately. The flexible membranes of these suction cups are pulled up by SMA wires and thus a vacuum is created between the membrane and the work piece surface. The self-sensing ability of the SMA wires are used in both prototypes for monitoring their actuation.

  5. A sensor-based automation system for handling nuclear materials

    International Nuclear Information System (INIS)

    An automated system is being developed for handling large payloads of radioactive nuclear materials in an analytical laboratory. The automation system performs unpacking and repacking of payloads from shipping and storage containers, and delivery of the payloads to the stations in the laboratory. The system uses machine vision and force/torque sensing to provide sensor-based control of the automation system in order to enhance system safety, flexibility, and robustness, and achieve easy remote operation. The automation system also controls the operation of the laboratory measurement systems and the coordination of them with the robotic system. Particular attention has been given to system design features and analytical methods that provide an enhanced level of operational safety. Independent mechanical gripper interlock and tool release mechanisms were designed to prevent payload mishandling. An extensive Failure Modes and Effects Analysis of the automation system was developed as a safety design analysis tool

  6. Staging queues in material handling and transportation systems

    OpenAIRE

    Gue, Kevin R.; Kang, Keebom

    2001-01-01

    Proceedings of the 2001 Winter Simulation Conference B. A. Peters, J. S. Smith, D. J. Medeiros, and M. W. Rohrer, eds. In most physical queueing applications, customers join a queue andmove forward after each service, leaving room for others to join behind them. Some queues found in material handling and transportation systems do not operate like this because the queued entities (pallets or unoccupied cars, for example) are incapable of moving forward autonomously. We ...

  7. System design for safe robotic handling of nuclear materials

    International Nuclear Information System (INIS)

    Robotic systems are being developed by the Intelligent Systems and Robotics Center at Sandia National Laboratories to perform automated handling tasks with radioactive nuclear materials. These systems will reduce the occupational radiation exposure to workers by automating operations which are currently performed manually. Because the robotic systems will handle material that is both hazardous and valuable, the safety of the operations is of utmost importance; assurance must be given that personnel will not be harmed and that the materials and environment will be protected. These safety requirements are met by designing safety features into the system using a layered approach. Several levels of mechanical, electrical and software safety prevent unsafe conditions from generating a hazard, and bring the system to a safe state should an unexpected situation arise. The system safety features include the use of industrial robot standards, commercial robot systems, commercial and custom tooling, mechanical safety interlocks, advanced sensor systems, control and configuration checks, and redundant control schemes. The effectiveness of the safety features in satisfying the safety requirements is verified using a Failure Modes and Effects Analysis. This technique can point out areas of weakness in the safety design as well as areas where unnecessary redundancy may reduce the system reliability

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

    International Nuclear Information System (INIS)

    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)

  9. Engineered Barrier System - Assessment of the Corrosion Properties of Copper Canisters. Report from a Workshop. Synthesis and extended abstract

    International Nuclear Information System (INIS)

    valid at some stage during the repository evolution. Workshop participants suggested a need for SKI to review SKB's canister corrosion model in more detail as part of future safety assessment reviews (calculations, assumptions and data). Additional experimental work might be needed for the assessment of copper corrosion in high chloride environments and with simultaneous presence of chloride and sulphide. It is essential that altogether consistent facts, understanding and models are used when developing an argument. Any inconsistency regarding these three aspects (facts, understanding, models) needs to be identified. An example would be if thermodynamic data and theoretical calculations suggest that corrosion will not happen, while kinetic data (experimental results) suggest a significant corrosion rate. For future safety assessments, SKB is recommended to use a consistent template for the handling of different corrosion mechanisms even if their final treatment will be quite different. This may include e.g. an extended application of the exclusion principle and/or application of the decision tree approach (as applied for stress corrosion cracking in the Canadian programme). However, it should be noted that the reliability of the exclusion principle depends on the quantity and quality of information on which it is based, and that more explicit criteria might be needed to support the decision tree approach. There is also a need for a well structured approach to handling uncertainties. Examples include those that can be characterised as variability (welding defects, sulphide content of groundwater and bentonite) and as lack of knowledge (e.g. microbial viability, the existence of an unidentified groundwater component affecting corrosion or an unknown corrosion mechanism). A suitable combination of a probabilistic application of the main copper corrosion model, well supported calculation cases with mechanistic models and possibly a selection of what-if calculations could

  10. Engineered Barrier System - Assessment of the Corrosion Properties of Copper Canisters. Report from a Workshop. Synthesis and extended abstract

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Peter (ed.) [Quintessa Ltd., Henley-on-Thames (GB)] (and others)

    2006-03-15

    assumption turns out not to be valid at some stage during the repository evolution. Workshop participants suggested a need for SKI to review SKB's canister corrosion model in more detail as part of future safety assessment reviews (calculations, assumptions and data). Additional experimental work might be needed for the assessment of copper corrosion in high chloride environments and with simultaneous presence of chloride and sulphide. It is essential that altogether consistent facts, understanding and models are used when developing an argument. Any inconsistency regarding these three aspects (facts, understanding, models) needs to be identified. An example would be if thermodynamic data and theoretical calculations suggest that corrosion will not happen, while kinetic data (experimental results) suggest a significant corrosion rate. For future safety assessments, SKB is recommended to use a consistent template for the handling of different corrosion mechanisms even if their final treatment will be quite different. This may include e.g. an extended application of the exclusion principle and/or application of the decision tree approach (as applied for stress corrosion cracking in the Canadian programme). However, it should be noted that the reliability of the exclusion principle depends on the quantity and quality of information on which it is based, and that more explicit criteria might be needed to support the decision tree approach. There is also a need for a well structured approach to handling uncertainties. Examples include those that can be characterised as variability (welding defects, sulphide content of groundwater and bentonite) and as lack of knowledge (e.g. microbial viability, the existence of an unidentified groundwater component affecting corrosion or an unknown corrosion mechanism). A suitable combination of a probabilistic application of the main copper corrosion model, well supported calculation cases with mechanistic models and possibly a selection

  11. Mechanical integrity of canisters

    International Nuclear Information System (INIS)

    This document constitutes the final report from 'SKBs reference group for mechanical integrity of canisters for spent nuclear fuel'. A complete list of all reports initiated by the reference group can be found in the summary report in this document. The main task of the reference group has been to advice SKB regarding the choice (ranking of alternatives) of canister type for different types of storage. The choice should be based on requirements of impermeability for a given time period and identification of possible limiting mechanisms. The main conclusions from the work were: From mechanical point of view, low phosphorous oxygen free copper (Cu-OFP) is a preferred canisters material. It exhibits satisfactory ductility both during tensile and creep testing. The residual stresses in the canisters are of such a magnitude that the estimated time to creep rupture with the data obtained for the Cu-OFP material is essentially infinite. Based on the present knowledge of stress corrosion cracking of copper there appears to be a small risk for such to occur in the projected environment. This risk need some further study. Rock shear movements of the size of 10 cm should pose no direct threat to the integrity of the canisters. Considering mechanical integrity, the composite copper/steel canister is an advantageous alternative. The recommendations for further research included continued studies of the creep properties of copper and of stress corrosion cracking. However, the studies should focus more directly on the design and fabrication aspect of the canister

  12. The concrete canister program

    International Nuclear Information System (INIS)

    In the spring of 1974, WNRE began development and demonstration of a dry storage concept, called the concrete canister, as a possible alternative to storage of irradiated CANDU fuel in water pools. The canister is a thick-walled concrete monolith containing baskets of fuel in the dry state. The decay heat from the fuel is dissipated to the environment by natural heat transfer. Four canisters were designed and constructed. Two canisters containing electric heaters have been subjected to heat loads of 2.5 times the design, ramp heat-load cycling, and simulated weathering tests. The other two canisters were loaded with irradiated fuel, one containing fuel bundles of uniform decay heat and the other containing bundles of non-uniform decay heat in a non-symmetrical radial and axial array. The collected data were used to verify the analytical tools for prediction of effectiveness of heat transfer and radiation shielding and to verify the design of the basket and canisters. The demonstration canisters have shown that this concept is a viable alternative to water pools for the storage of irradiated CANDU fuel. (author)

  13. Three-dimensional television system for remote handling

    International Nuclear Information System (INIS)

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

  14. Analyzing Graph Transformation Systems through Constraint Handling Rules

    OpenAIRE

    Raiser, Frank; Frühwirth, Thom

    2010-01-01

    Graph transformation systems (GTS) and constraint handling rules (CHR) are non-deterministic rule-based state transition systems. CHR is well-known for its powerful confluence and program equivalence analyses, for which we provide the basis in this work to apply them to GTS. We give a sound and complete embedding of GTS in CHR, investigate confluence of an embedded GTS, and provide a program equivalence analysis for GTS via the embedding. The results confirm the suitability of CHR-based progr...

  15. Pneumatic Conveying System For Chilli Handling: A Review

    OpenAIRE

    J. M. Mahure; P. G. Mehar; S. R. Ikhar; A. V. Vanalkar

    2013-01-01

    Abstract:In the industries where bulk material is to be transferred from one place in the process plant to the other, material handling systems are required. Various types of conveyors are available in the market having their own characteristic features. But in the industries where very high mass flow rates are required, pneumatic conveying system can be very useful. Various other conveyors are also present but some occupy a lot of space in the plant whereas some cannot give such high mass fl...

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

    OpenAIRE

    Franck Dechelette; Franck Morin; Guy Laffont; Gilles Rodriguez; Emmanuel Sanseigne; Sébastien Christin; Xavier Mognot; Aurélien Morcillo

    2014-01-01

    International audience 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...

  17. Alarm handling systems and techniques developed to match operator tasks

    International Nuclear Information System (INIS)

    This paper covers alarm handling methods and techniques explored at the Halden Project, and describes current status on the research activities on alarm systems. Alarm systems are often designed by application of a bottom-up strategy, generating alarms at component level. If no structuring of the alarms is applied, this may result in alarm avalanches in major plant disturbances, causing cognitive overload of the operator. An alarm structuring module should be designed using a top-down approach, analysing operator's tasks, plant states, events and disturbances. One of the operator's main tasks during plant disturbances is status identification, including determination of plant status and detection of plant anomalies. The main support of this is provided through the alarm systems, the process formats, the trends and possible diagnosis systems. The alarm system should both physically and conceptually be integrated with all these systems. 9 refs, 5 figs

  18. Fuel handling control system for TAPP-3,4

    International Nuclear Information System (INIS)

    The fuel handling system is required for loading of new fuel bundles in the reactor, to retrieve spent fuel bundles and move them to the storage bay during online condition. The Fuel Handling operations involve a large number of steps that are to be executed automatically in a specified and sequential manner. Each step includes checking of permissive and issuing the required commands. The commands effect the selection of process set points, accurate positioning of mechanical rams at pre-calibrated discrete positions, advancing/ retracting of hydraulic cylinders, opening/ closing of valves, energizing or de-energizing of relays and provide adequate indications or messages. There are large number of inputs and outputs in the control system and an extensive logic relates the output to input. The system provides ON-POWER refueling at any selected reactor coolant channel; the operation is completely synchronized between two sides fueling machines. A comprehensive status of system is presented on OIS. It also permits manual interventions to complete the execution of current step or continue sequence operation. The overall design of the control system is such as to ensure safety under various conditions of postulated failures. (author)

  19. Transportation system (TRUPACT) for contact-handled transuranic wastes

    International Nuclear Information System (INIS)

    Contact-handled transuranic defense waste is being, and will continue to be, moved between a number of locations in the United States. The DOE is sponsoring development of safe, efficient, licensable, and cost-effective transportation systems to handle this waste. The systems being developed have been named TRUPACT which stands for TRansUranic PACkage Transporter. The system will be compatible with Type A packagings used by waste generators, interim storage facilities, and repositories. TRUPACT is required to be a Type B packaging since larger than Type A quantities of some radionuclides (particularly plutonium) may be involved in the collection of Type A packagings. TRUPACT must provide structural and thermal protection to the waste in hypothetical accident environments specified in DOT regulations 49CFR173 and NRC regulations 10CFR71. Preliminary design of the systems has been completed and final design for a truck system is underway. The status of the development program is reviewed in this paper and the reference design is described. Tests that have been conducted are discussed and long-term program objectives are reviewed

  20. EVALUATION OF REQUIREMENTS FOR THE DWPF HIGHER CAPACITY CANISTER

    Energy Technology Data Exchange (ETDEWEB)

    Miller, D.; Estochen, E.; Jordan, J.; Kesterson, M.; Mckeel, C.

    2014-08-05

    The Defense Waste Processing Facility (DWPF) is considering the option to increase canister glass capacity by reducing the wall thickness of the current production canister. This design has been designated as the DWPF Higher Capacity Canister (HCC). A significant decrease in the number of canisters processed during the life of the facility would be achieved if the HCC were implemented leading to a reduced overall reduction in life cycle costs. Prior to implementation of the change, Savannah River National Laboratory (SRNL) was requested to conduct an evaluation of the potential impacts. The specific areas of interest included loading and deformation of the canister during the filling process. Additionally, the effect of the reduced wall thickness on corrosion and material compatibility needed to be addressed. Finally the integrity of the canister during decontamination and other handling steps needed to be determined. The initial request regarding canister fabrication was later addressed in an alternate study. A preliminary review of canister requirements and previous testing was conducted prior to determining the testing approach. Thermal and stress models were developed to predict the forces on the canister during the pouring and cooling process. The thermal model shows the HCC increasing and decreasing in temperature at a slightly faster rate than the original. The HCC is shown to have a 3°F ΔT between the internal and outer surfaces versus a 5°F ΔT for the original design. The stress model indicates strain values ranging from 1.9% to 2.9% for the standard canister and 2.5% to 3.1% for the HCC. These values are dependent on the glass level relative to the thickness transition between the top head and the canister wall. This information, along with field readings, was used to set up environmental test conditions for corrosion studies. Small 304-L canisters were filled with glass and subjected to accelerated environmental testing for 3 months. No evidence of

  1. Progress in standardization for ITER Remote Handling control system

    International Nuclear Information System (INIS)

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

  2. Design considerations for laboratory robotics systems handling toxic substances

    International Nuclear Information System (INIS)

    This paper discusses design issues pertinent to development of reliable robotic systems for handling toxic substances, using examples from Midwest Research Institute's experience with the full range of materials selected by the federal government for toxicology testing. This includes laboratory work with substances known to have adverse toxicological effects (e.g., known carcinogens) as well as chemicals selected for evaluation of their toxic potential. Chemicals under evaluation include pharmaceuticals, industrial chemicals, solvents, pesticides, and a variety of other chemicals, solvents, pesticides, and a variety of other chemicals to which the populace is exposed in the workplace or in their daily lives

  3. Drop Calculations of HLW Canister and Pu Can-in-Canister

    International Nuclear Information System (INIS)

    The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document

  4. Drop Calculations of HLW Canister and Pu Can-in-Canister

    Energy Technology Data Exchange (ETDEWEB)

    Sreten Mastilovic

    2001-07-31

    The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C ), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document.

  5. Reliability assessment of waste-handling-building HVAC system

    International Nuclear Information System (INIS)

    This paper presents a method to estimate the probability of an unfiltered release to the environment due to failure of the heating, ventilation, and air-conditioning (HVAC) system in the Waste Handling Building (WHB) at the proposed waste repository at Yucca Mountain. The scope of the calculation is limited to the function of the WHB HVAC system to maintain its once-through capability, that is, the ability to draw air through the various confinement areas through the high-efficiency particulate air (HEPA) filters, before exhausting the air to the atmosphere through the exhaust air stack. The ability to draw air through the WHB also maintains negative pressure between confinement areas (causing air flow inward rather than toward the environment). Other functions of the HVAC system, e.g., room heating and cooling for comfort, were not considered in this analysis

  6. Handling encapsulated spent fuel in a geologic repository environment

    International Nuclear Information System (INIS)

    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

  7. Canister Transfer Facility Criticality Calculations

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Monroe-Rammsy

    2000-10-13

    The objective of this calculation is to evaluate the criticality risk in the surface facility for design basis events (DBE) involving Department of Energy (DOE) Spent Nuclear Fuel (SNF) standardized canisters (Civilian Radioactive Waste Management System [CRWMS] Management and Operating Contractor [M&O] 2000a). Since some of the canisters will be stored in the surface facility before they are loaded in the waste package (WP), this calculation supports the demonstration of concept viability related to the Surface Facility environment. The scope of this calculation is limited to the consideration of three DOE SNF fuels, specifically Enrico Fermi SNF, Training Research Isotope General Atomic (TRIGA) SNF, and Mixed Oxide (MOX) Fast Flux Test Facility (FFTF) SNF.

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

    International Nuclear Information System (INIS)

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

  9. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel (SNF) Canister Storage Building (CSB) is the interim storage facility for the K-Basin SNF at the US. Department of Energy (DOE) Hanford Site. The SNF is packaged in multi-canister overpacks (MCOs). The MCOs are placed inside transport casks, then delivered to the service station inside the CSB. At the service station, the MCO handling machine (MHM) moves the MCO from the cask to a storage tube or one of two sample/weld stations. There are 220 standard storage tubes and six overpack storage tubes in a below grade reinforced concrete vault. Each storage tube can hold two MCOs

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

    International Nuclear Information System (INIS)

    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)

  11. Design and full scale test of the fuel handling system

    International Nuclear Information System (INIS)

    In the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) fuel elements move through the core driven by gravity. To reach their design burn-up the fuel elements are re-shuttled five times. This transportation outside the core is mainly achieved pneumatically. Although, adopting the international experience at design and operation of similar systems some key components were improved so that the fuel handling system (FHS) becomes simpler and more reliable. The improved components were tested in full-scale testing facilities. The debugging test and the first loading operation for the FHS indicate that the FHS meets the demands of the HTR-10. In this paper, the functions, design parameters, technological processes, main components and design characteristics of the FHS are described in detail. The flow schemes, design parameters of the full-scale testing facilities and the experimental results are briefly introduced

  12. A scintillator purification plant and fluid handling system for SNO+

    Energy Technology Data Exchange (ETDEWEB)

    Ford, Richard J., E-mail: ford@snolab.ca [SNOLAB, Creighton Mine #9, 1039 R.R.24, Lively, Ontario, Canada. (Canada)

    2015-08-17

    A large capacity purification plant and fluid handling system has been constructed for the SNO+ neutrino and double-beta decay experiment, located 6800 feet underground at SNOLAB, Canada. SNO+ is a refurbishment of the SNO detector to fill the acrylic vessel with liquid scintillator based on Linear Alkylbenzene (LAB) and 2 g/L PPO, and also has a phase to load natural tellurium into the scintillator for a double-beta decay experiment with {sup 130}Te. The plant includes processes multi-stage dual-stream distillation, column water extraction, steam stripping, and functionalized silica gel adsorption columns. The plant also includes systems for preparing the scintillator with PPO and metal-loading the scintillator for double-beta decay exposure. We review the basis of design, the purification principles, specifications for the plant, and the construction and installations. The construction and commissioning status is updated.

  13. A scintillator purification plant and fluid handling system for SNO+

    International Nuclear Information System (INIS)

    A large capacity purification plant and fluid handling system has been constructed for the SNO+ neutrino and double-beta decay experiment, located 6800 feet underground at SNOLAB, Canada. SNO+ is a refurbishment of the SNO detector to fill the acrylic vessel with liquid scintillator based on Linear Alkylbenzene (LAB) and 2 g/L PPO, and also has a phase to load natural tellurium into the scintillator for a double-beta decay experiment with 130Te. The plant includes processes multi-stage dual-stream distillation, column water extraction, steam stripping, and functionalized silica gel adsorption columns. The plant also includes systems for preparing the scintillator with PPO and metal-loading the scintillator for double-beta decay exposure. We review the basis of design, the purification principles, specifications for the plant, and the construction and installations. The construction and commissioning status is updated

  14. Handling Overload Conditions In High Performance Trustworthy Information Retrieval Systems

    CERN Document Server

    Ramachandran, Sumalatha; Paulraj, Sujaya; Ramaraj, Vetriselvi

    2010-01-01

    Web search engines retrieve a vast amount of information for a given search query. But the user needs only trustworthy and high-quality information from this vast retrieved data. The response time of the search engine must be a minimum value in order to satisfy the user. An optimum level of response time should be maintained even when the system is overloaded. This paper proposes an optimal Load Shedding algorithm which is used to handle overload conditions in real-time data stream applications and is adapted to the Information Retrieval System of a web search engine. Experiment results show that the proposed algorithm enables a web search engine to provide trustworthy search results to the user within an optimum response time, even during overload conditions.

  15. A Scintillator Purification Plant and Fluid Handling System for SNO+

    CERN Document Server

    Ford, Richard J

    2015-01-01

    A large capacity purification plant and fluid handling system has been constructed for the SNO+ neutrino and double-beta decay experiment, located 6800 feet underground at SNOLAB, Canada. SNO+ is a refurbishment of the SNO detector to fill the acrylic vessel with liquid scintillator based on Linear Alkylbenzene (LAB) and 2 g/L PPO, and also has a phase to load natural tellurium into the scintillator for a double-beta decay experiment with 130Te. The plant includes processes multi-stage dual-stream distillation, column water extraction, steam stripping, and functionalized silica gel adsorption columns. The plant also includes systems for preparing the scintillator with PPO and metal-loading the scintillator for double-beta decay exposure. We review the basis of design, the purification principles, specifications for the plant, and the construction and installations. The construction and commissioning status is updated.

  16. Integrated Payload Data Handling Systems Using Software Partitioning

    Science.gov (United States)

    Taylor, Alun; Hann, Mark; Wishart, Alex

    2015-09-01

    An integrated Payload Data Handling System (I-PDHS) is one in which multiple instruments share a central payload processor for their on-board data processing tasks. This offers a number of advantages over the conventional decentralised architecture. Savings in payload mass and power can be realised because the total processing resource is matched to the requirements, as opposed to the decentralised architecture here the processing resource is in effect the sum of all the applications. Overall development cost can be reduced using a common processor. At individual instrument level the potential benefits include a standardised application development environment, and the opportunity to run the instrument data handling application on a fully redundant and more powerful processing platform [1]. This paper describes a joint program by SCISYS UK Limited, Airbus Defence and Space, Imperial College London and RAL Space to implement a realistic demonstration of an I-PDHS using engineering models of flight instruments (a magnetometer and camera) and a laboratory demonstrator of a central payload processor which is functionally representative of a flight design. The objective is to raise the Technology Readiness Level of the centralised data processing technique by address the key areas of task partitioning to prevent fault propagation and the use of a common development process for the instrument applications. The project is supported by a UK Space Agency grant awarded under the National Space Technology Program SpaceCITI scheme. [1].

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

    International Nuclear Information System (INIS)

    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

  18. Developing a structural health monitoring system for nuclear dry cask storage canister

    Science.gov (United States)

    Sun, Xiaoyi; Lin, Bin; Bao, Jingjing; Giurgiutiu, Victor; Knight, Travis; Lam, Poh-Sang; Yu, Lingyu

    2015-03-01

    Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. In total, there are over 1482 dry cask storage system (DCSS) in use at US plants, storing 57,807 fuel assemblies. Nondestructive material condition monitoring is in urgent need and must be integrated into the fuel cycle to quantify the "state of health", and more importantly, to guarantee the safe operation of radioactive waste storage systems (RWSS) during their extended usage period. A state-of-the-art nuclear structural health monitoring (N-SHM) system based on in-situ sensing technologies that monitor material degradation and aging for nuclear spent fuel DCSS and similar structures is being developed. The N-SHM technology uses permanently installed low-profile piezoelectric wafer sensors to perform long-term health monitoring by strategically using a combined impedance (EMIS), acoustic emission (AE), and guided ultrasonic wave (GUW) approach, called "multimode sensing", which is conducted by the same network of installed sensors activated in a variety of ways. The system will detect AE events resulting from crack (case for study in this project) and evaluate the damage evolution; when significant AE is detected, the sensor network will switch to the GUW mode to perform damage localization, and quantification as well as probe "hot spots" that are prone to damage for material degradation evaluation using EMIS approach. The N-SHM is expected to eventually provide a systematic methodology for assessing and monitoring nuclear waste storage systems without incurring human radiation exposure.

  19. Feasibility study for a DOE research and production fuel multipurpose canister

    International Nuclear Information System (INIS)

    This is a report of the feasibility of multipurpose canisters for transporting, storing, and sing of Department of Energy research and production spent nuclear fuel. Six representative Department of Energy fuel assemblies were selected, and preconceptual canister designs were developed to accommodate these assemblies. The study considered physical interface, structural adequacy, criticality safety, shielding capability, thermal performance of the canisters, and fuel storage site infrastructure. The external envelope of the canisters was designed to fit within the overpack casks for commercial canisters being developed for the Department of Energy Office of Civilian Radioactive Waste Management. The budgetary cost of canisters to handle all fuel considered is estimated at $170.8M. One large conceptual boiling water reactor canister design, developed for the Office of Civilian Radioactive Waste Management, and two new canister designs can accommodate at least 85% of the volume of the Department of Energy fuel considered. Canister use minimizes public radiation exposure and is cost effective compared with bare fuel handling. Results suggest the need for additional study of issues affecting canister use and for conceptual design development of the three canisters

  20. A high intensity beam handling system at the KEK-PS new experimental hall

    International Nuclear Information System (INIS)

    We would like to summarize newly developed technology for handling high-intensity beams. This was practically employed in the beam-handling system of primary protons at the KEK-PS new experimental hall. (author)

  1. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

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

  2. Improved Design of EURISOL Fission Target Systems and Handling

    CERN Document Server

    F. Negoita, L. Serbina, E. Udup, O. Alyakrinskiy, M. Barbui, J. Bermudez, L.B. Tecchio, Y. Kadi, C. Kharoua, Y. Romanets

    The report discusses the constraints, solutions and options for integration of the main systems needed to assure the operation and maintenance/exchange of the fission target – ion sources assemblies capable to withstand very high neutron fluxes around EURISOL multi-MW neutron converter, to provide the desired fission rate of 10e15 fis./sec. and to extract with highest possible efficiencies the fission products as radioactive ion beam. Adapting the concepts developed for a single fission target within PIAFE and MAFF projects, simultaneous operation of six fission targets was proposed for EURISOL. A first design following these concepts was presented in a previous report. Here we describe several improvements implemented in the design for a more efficient, more reliable and safer operation. An optimized layout of the part of the EURISOL MMW station related to fission target services and handling is proposed as well.

  3. 324 Building liquid waste handling and removal system project plan

    International Nuclear Information System (INIS)

    This report evaluates the modification options for handling radiological liquid waste generated during decontamination and cleanout of the 324 Building. Recent discussions indicate that the Hanford site railroad system will be closed by the end of FY 1998 necessitating the need for an alternate transfer method. The issue of handling of Radioactive Liquid Waste (RLW) from the 324 Building (assuming the 340 Facility is not available to accept the RLW) has been examined in at least two earlier engineering studies (Parsons 1997a and Hobart 1997). Each study identified a similar preferred alternative that included modifying the 324 Building RLWS to allow load-out of wastewater to a truck tanker, while making maximum use of existing piping, tanks, instrumentation, controls and other features to minimize costs and physical changes to the building. This alternative is accepted as the basis for further discussion presented in this study. The goal of this engineering study is to verify the path forward presented in the previous studies and assure that the selected alternative satisfies the 324 Building deactivation goals and objectives as currently described in the project management plan. This study will also evaluate options available to implement the preferred alternative and select the preferred option for implementation of the entire system. Items requiring further examination will also be identified. Finally, the study will provide a conceptual design, schedule and cost estimate for the required modifications to the 324 Building to allow removal of RLW. Attachment 5 is an excerpt from the project baseline schedule found in the Project Management Plan

  4. POD (Probability of Detection) evaluation of NDT techniques for Cu-canisters for risk assessment of nuclear waste encapsulation

    International Nuclear Information System (INIS)

    In order to handle long living radioactive waste Sweden is planning to build a deep repository that requires no monitoring by future generations. The spent nuclear fuel will be encapsulated in copper canisters consisting of a graphite cast iron insert shielded by an outer 30-50 mm thick copper cylinder for corrosion and radiation protection. The cast iron insert provides the necessary strength and shielding of radiation. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel at the canister lab in Oskarshamn, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). Mid 2005 SKB decided that FSW is the preferred sealing technique. A subpart of the final risk assessment for the deep repository is to determine the risk of premature canister leak caused by discontinuities in the insert, in the sealing weld or elsewhere in the copper shielding. Therefore the quality of the production processes and the reliability of the NDT system must be satisfactorily determined and combined to derive assumptions regarding the frequency of undetected production discontinuities in relation to the acceptance criteria for the ensemble of canisters. The reliability of the NDT systems can be derived from POD curves which are investigated for X-ray and ultrasonic techniques applied by SKB. The POD evaluation was carried out by BAM in a joint project for SKB and is evaluated within the common ''a versus a'' approach according to the MIL1823 and some extensions due to the more complex flaw situations in the canisters compared to the original aerospace applications. (orig.)

  5. Mark 4A antenna control system data handling architecture study

    Science.gov (United States)

    Briggs, H. C.; Eldred, D. B.

    1991-01-01

    A high-level review was conducted to provide an analysis of the existing architecture used to handle data and implement control algorithms for NASA's Deep Space Network (DSN) antennas and to make system-level recommendations for improving this architecture so that the DSN antennas can support the ever-tightening requirements of the next decade and beyond. It was found that the existing system is seriously overloaded, with processor utilization approaching 100 percent. A number of factors contribute to this overloading, including dated hardware, inefficient software, and a message-passing strategy that depends on serial connections between machines. At the same time, the system has shortcomings and idiosyncrasies that require extensive human intervention. A custom operating system kernel and an obscure programming language exacerbate the problems and should be modernized. A new architecture is presented that addresses these and other issues. Key features of the new architecture include a simplified message passing hierarchy that utilizes a high-speed local area network, redesign of particular processing function algorithms, consolidation of functions, and implementation of the architecture in modern hardware and software using mainstream computer languages and operating systems. The system would also allow incremental hardware improvements as better and faster hardware for such systems becomes available, and costs could potentially be low enough that redundancy would be provided economically. Such a system could support DSN requirements for the foreseeable future, though thorough consideration must be given to hard computational requirements, porting existing software functionality to the new system, and issues of fault tolerance and recovery.

  6. The functioning capability of a mobile remote handling system

    International Nuclear Information System (INIS)

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

  7. Radial Internal Material Handling System (RIMS) for Circular Habitat Volumes

    Science.gov (United States)

    Howe, Alan S.; Haselschwardt, Sally; Bogatko, Alex; Humphrey, Brian; Patel, Amit

    2013-01-01

    On planetary surfaces, pressurized human habitable volumes will require a means to carry equipment around within the volume of the habitat, regardless of the partial gravity (Earth, Moon, Mars, etc.). On the NASA Habitat Demonstration Unit (HDU), a vertical cylindrical volume, it was determined that a variety of heavy items would need to be carried back and forth from deployed locations to the General Maintenance Work Station (GMWS) when in need of repair, and other equipment may need to be carried inside for repairs, such as rover parts and other external equipment. The vertical cylindrical volume of the HDU lent itself to a circular overhead track and hoist system that allows lifting of heavy objects from anywhere in the habitat to any other point in the habitat interior. In addition, the system is able to hand-off lifted items to other material handling systems through the side hatches, such as through an airlock. The overhead system consists of two concentric circle tracks that have a movable beam between them. The beam has a hoist carriage that can move back and forth on the beam. Therefore, the entire system acts like a bridge crane curved around to meet itself in a circle. The novelty of the system is in its configuration, and how it interfaces with the volume of the HDU habitat. Similar to how a bridge crane allows coverage for an entire rectangular volume, the RIMS system covers a circular volume. The RIMS system is the first generation of what may be applied to future planetary surface vertical cylinder habitats on the Moon or on Mars.

  8. Evolution of a test article handling system for the SP-100 ground engineering system test

    International Nuclear Information System (INIS)

    A simulated space environment test of a flight prototypic SP-100 reactor, control system, and flight shield will be conducted at the Hanford Engineering Development Laboratory (HEDL). The flight prototypic components and the supporting primary heat removal system are collectively known as the Nuclear Assembly Test Article (TA). The unique configuration and materials of fabrication for the Test Article require a specialized handling facility to support installation, maintenance, and final disposal operations. Westinghouse Hanford Company, the Test Site Operator, working in conjunction with General Electric Company, the Test Article supplier, developed and evaluated several handling concepts resulting in the selection of a reference Test Article Handling System. The development of the reference concept for the handling system is presented

  9. Quinone-induced protein handling changes: Implications for major protein handling systems in quinone-mediated toxicity

    Energy Technology Data Exchange (ETDEWEB)

    Xiong, Rui; Siegel, David; Ross, David, E-mail: david.ross@ucdenver.edu

    2014-10-15

    Para-quinones such as 1,4-Benzoquinone (BQ) and menadione (MD) and ortho-quinones including the oxidation products of catecholamines, are derived from xenobiotics as well as endogenous molecules. The effects of quinones on major protein handling systems in cells; the 20/26S proteasome, the ER stress response, autophagy, chaperone proteins and aggresome formation, have not been investigated in a systematic manner. Both BQ and aminochrome (AC) inhibited proteasomal activity and activated the ER stress response and autophagy in rat dopaminergic N27 cells. AC also induced aggresome formation while MD had little effect on any protein handling systems in N27 cells. The effect of NQO1 on quinone induced protein handling changes and toxicity was examined using N27 cells stably transfected with NQO1 to generate an isogenic NQO1-overexpressing line. NQO1 protected against BQ–induced apoptosis but led to a potentiation of AC- and MD-induced apoptosis. Modulation of quinone-induced apoptosis in N27 and NQO1-overexpressing cells correlated only with changes in the ER stress response and not with changes in other protein handling systems. These data suggested that NQO1 modulated the ER stress response to potentiate toxicity of AC and MD, but protected against BQ toxicity. We further demonstrated that NQO1 mediated reduction to unstable hydroquinones and subsequent redox cycling was important for the activation of the ER stress response and toxicity for both AC and MD. In summary, our data demonstrate that quinone-specific changes in protein handling are evident in N27 cells and the induction of the ER stress response is associated with quinone-mediated toxicity. - Highlights: • Unstable hydroquinones contributed to quinone-induced ER stress and toxicity.

  10. Quinone-induced protein handling changes: Implications for major protein handling systems in quinone-mediated toxicity

    International Nuclear Information System (INIS)

    Para-quinones such as 1,4-Benzoquinone (BQ) and menadione (MD) and ortho-quinones including the oxidation products of catecholamines, are derived from xenobiotics as well as endogenous molecules. The effects of quinones on major protein handling systems in cells; the 20/26S proteasome, the ER stress response, autophagy, chaperone proteins and aggresome formation, have not been investigated in a systematic manner. Both BQ and aminochrome (AC) inhibited proteasomal activity and activated the ER stress response and autophagy in rat dopaminergic N27 cells. AC also induced aggresome formation while MD had little effect on any protein handling systems in N27 cells. The effect of NQO1 on quinone induced protein handling changes and toxicity was examined using N27 cells stably transfected with NQO1 to generate an isogenic NQO1-overexpressing line. NQO1 protected against BQ–induced apoptosis but led to a potentiation of AC- and MD-induced apoptosis. Modulation of quinone-induced apoptosis in N27 and NQO1-overexpressing cells correlated only with changes in the ER stress response and not with changes in other protein handling systems. These data suggested that NQO1 modulated the ER stress response to potentiate toxicity of AC and MD, but protected against BQ toxicity. We further demonstrated that NQO1 mediated reduction to unstable hydroquinones and subsequent redox cycling was important for the activation of the ER stress response and toxicity for both AC and MD. In summary, our data demonstrate that quinone-specific changes in protein handling are evident in N27 cells and the induction of the ER stress response is associated with quinone-mediated toxicity. - Highlights: • Unstable hydroquinones contributed to quinone-induced ER stress and toxicity

  11. Integrated robotic vehicle control system for outdoor container handling

    Science.gov (United States)

    Viitanen, Jouko O.; Haverinen, Janne; Mattila, Pentti; Maekelae, Hannu; von Numers, Thomas; Stanek, Zbigniev; Roening, Juha

    1997-09-01

    We describe an integrated system developed for use onboard a moving work machine. The machine is targeted to such applications as e.g. automatic container handling at loading terminals. The main emphasis is on the various environment perception duties required by autonomous or semi-autonomous operation. These include obstacle detection, container position determination, localization needed for efficient navigation and measurement of docking and grasping locations of containers. Practical experience is reported on the use of several different types of technologies for the tasks. For close distance measurement, such as container row following, ultrasonic measurement was used, with associated control software. For obstacle and docking position detection, 3D active vision techniques were developed with structured lighting, utilizing also motion estimation techniques. Depth from defocus-based methods were developed for passive 3D vision. For localization, fusion of data from several sources was carried out. These included dead-reckoning data from odometry, an inertial unit, and several alternative external localization devices, i.e. real-time kinematic GPS, inductive and optical transponders. The system was integrated to run on a real-time operating system platform, using a high-level software specification tool that created the hierarchical control structure of the software.

  12. Improved Air-Treatment Canister

    Science.gov (United States)

    Boehm, A. M.

    1982-01-01

    Proposed air-treatment canister integrates a heater-in-tube water evaporator into canister header. Improved design prevents water from condensing and contaminating chemicals that regenerate the air. Heater is evenly spiraled about the inlet header on the canister. Evaporator is brazed to the header.

  13. Shaft shock absorber for a spent fuel canister

    International Nuclear Information System (INIS)

    The disposal canister for spent nuclear fuel will be transferred by a lift to the repository which is 500 m deep in the bedrock. Model tests were carried out with an objective to estimate weather feasible shock absorber can be developed against the design accident case where the canister should survive a free fall to the lift shaft. If the velocity of the canister is not controlled by air drag or by any other deceleration means, the impact velocity may reach ultimate speed of 100 m/s. The canister would retain its integrity in impact on water when the bottom pit of the lift well is filled with groundwater. However, the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20 m high filling to the bottom pit of the lift well by Light Expanded Clay Aggregate (LECA), gives fair impact absorption to protect the fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (author)

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

    International Nuclear Information System (INIS)

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

  15. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    The Waste Acceptance Preliminary Specifications (WAPS) require that the contents of the canistered waste form are compatible with one another and the stainless steel canister. The canistered waste form is a closed system comprised of a stainless steel vessel containing waste glass, air, and condensate. This system will experience a radiation field and an elevated temperature due to radionuclide decay. This report discusses possible chemical reactions, radiation interactions, and corrosive reactions within this system both under normal storage conditions and after exposure to temperatures up to the normal glass transition temperature, which for DWPF waste glass will be between 440 and 460 degrees C. Specific conclusions regarding reactions and corrosion are provided. This document is based on the assumption that the period of interim storage prior to packaging at the federal repository may be as long as 50 years

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

    International Nuclear Information System (INIS)

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

  17. Dust removal experiments for ITER blanket remote handling system

    International Nuclear Information System (INIS)

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

  18. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

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

  19. Fuel handling and storage systems in nuclear power plants

    International Nuclear Information System (INIS)

    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

  20. Plutonium Can-In-Canister-Design Basis Event Analysis

    International Nuclear Information System (INIS)

    The purpose of this document is to perform a preliminary design basis event (DBE) analysis of the immobilized plutonium (can-in-canister) waste form to be referred to in this analysis as high level waste/plutonium (HLW/Pu). The objective of the analysis is to determine any preclosure safety impacts of the waste form on the Monitored Geologic Repository (MGR). The scope of this analysis is to determine the offsite dose consequences and associated frequencies of selected DBEs for systems handling disposable canisters that bound all surface and subsurface off-normal events, and to compare these results against regulatory limits. The results of this work are preliminary and are intended to be used to establish a set of preliminary MGR and waste form requirements, to identify mitigation or prevention options that may be required to meet regulatory limits, and to provide input to the Site Recommendation (SR) report. This document is prepared in accordance with the associated development plan (Civilian Radioactive Waste Management System Management and Operating Contractor [CRWMS M and O] 1999e)

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

    International Nuclear Information System (INIS)

    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)

  2. Status report, canister fabrication

    International Nuclear Information System (INIS)

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  3. Status report, canister fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Emilsson, Goeran [CSM Materialteknik AB, Linkoeping (Sweden)

    2004-06-01

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  4. Final Report: Part 1. In-Place Filter Testing Instrument for Nuclear Material Containers. Part 2. Canister Filter Test Standards for Aerosol Capture Rates.

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Austin Douglas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Runnels, Joel T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reeves, Kirk Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-11-02

    A portable instrument has been developed to assess the functionality of filter sand o-rings on nuclear material storage canisters, without requiring removal of the canister lid. Additionally, a set of fifteen filter standards were procured for verifying aerosol leakage and pressure drop measurements in the Los Alamos Filter Test System. The US Department of Energy uses several thousand canisters for storing nuclear material in different chemical and physical forms. Specialized filters are installed into canister lids to allow gases to escape, and to maintain an internal ambient pressure while containing radioactive contaminants. Diagnosing the condition of container filters and canister integrity is important to ensure worker and public safety and for determining the handling requirements of legacy apparatus. This report describes the In-Place-Filter-Tester, the Instrument Development Plan and the Instrument Operating Method that were developed at the Los Alamos National Laboratory to determine the “as found” condition of unopened storage canisters. The Instrument Operating Method provides instructions for future evaluations of as-found canisters packaged with nuclear material. Customized stainless steel canister interfaces were developed for pressure-port access and to apply a suction clamping force for the interface. These are compatible with selected Hagan-style and SAVY-4000 storage canisters that were purchased from NFT (Nuclear Filter Technology, Golden, CO). Two instruments were developed for this effort: an initial Los Alamos POC (Proof-of-Concept) unit and the final Los Alamos IPFT system. The Los Alamos POC was used to create the Instrument Development Plan: (1) to determine the air flow and pressure characteristics associated with canister filter clogging, and (2) to test simulated configurations that mimicked canister leakage paths. The canister leakage scenarios included quantifying: (A) air leakage due to foreign material (i.e. dust and hair

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

  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. K West Basin canister survey

    International Nuclear Information System (INIS)

    A survey was conducted of the K West Basin to determine the distribution of canister types that contain the irradiated N Reactor fuel. An underwater camera was used to conduct the survey during June 1998, and the results were recorded on videotape. A full row-by-row survey of the entire basin was performed, with the distinction between aluminum and stainless steel Mark 1 canisters made by the presence or absence of steel rings on the canister trunions (aluminum canisters have the steel rings). The results of the survey are presented in tables and figures. Grid maps of the three bays show the canister lid ID number and the canister type in each location that contained fuel. The following abbreviations are used in the grid maps for canister type designation: IA = Mark 1 aluminum, IS = Mark 1 stainless steel, and 2 = Mark 2 stainless steel. An overall summary of the canister distribution survey is presented in Table 1. The total number of canisters found to contain fuel was 3842, with 20% being Mark 1 Al, 25% being Mark 1 SS, and 55% being Mark 2 SS. The aluminum canisters were predominantly located in the East and West bays of the basin

  9. STS-100 MPLM Raffaello is moved to the payload canister

    Science.gov (United States)

    2001-01-01

    KENNEDY SPACE CENTER, Fla. - Workers inside the payload canister wait for the Multi-Purpose Logistics Module Raffaello to be lowered inside. It joins the Canadian robotic arm, SSRMS, already in place. Both elements are part of the payload on mission STS- 100 to the International Space Station. Raffaello carries six system racks and two storage racks for the U.S. Lab. The arm has seven motorized joints and is capable of handling large payloads and assisting with docking the Space Shuttle. The SSRMS is self- relocatable with a Latching End Effector so it can be attached to complementary ports spread throughout the Station'''s exterior surfaces. Launch of STS-100 is scheduled for April 19, 2001 at 2:41 p.m. EDT from Launch Pad 39A.

  10. Desludging of N Reactor fuel canisters: Analysis, Test, and data requirements

    International Nuclear Information System (INIS)

    The N Reactor fuel is currently stored in canisters in the K East (KE) and K West (KW) Basins. In KE, the canisters have open tops; in KW, the cans have sealed lids, but are vented to release gases. Corrosion products have formed on exposed uranium metal fuel, on carbon steel basin component surfaces, and on aluminum alloy canister surfaces. Much of the corrosion product is retained on the corroding surfaces; however, large inventories of particulates have been released. Some of the corrosion product particulates form sludge on the basin floors; some particulates are retained within the canisters. The floor sludge inventories are much greater in the KE Basin than in the KW Basin because KE Basin operated longer and its water chemistry was less controlled. Another important factor is the absence of lids on the KE canisters, allowing uranium corrosion products to escape and water-borne species, principally iron oxides, to settle in the canisters. The inventories of corrosion products, including those released as particulates inside the canisters, are only beginning to be characterized for the closed canisters in KW Basin. The dominant species in the KE floor sludge are oxides of aluminum, iron, and uranium. A large fraction of the aluminum and uranium floor sludge particulates may have been released during a major fuel segregation campaign in the 1980s, when fuel was emptied from 4990 canisters. Handling and jarring of the fuel and aluminum canisters seems likely to have released particulates from the heavily corroded surfaces. Four candidate methods are discussed for dealing with canister sludge emerged in the N Reactor fuel path forward: place fuel in multi-canister overpacks (MCOs) without desludging; drill holes in canisters and drain; drill holes in canisters and flush with water; and remove sludge and repackage the fuel

  11. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    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.)

  12. STUDY OF A PUBLIC STORAGE AND HANDLING SYSTEM OF MEDICINES

    Directory of Open Access Journals (Sweden)

    KEILA RAQUEL DOS SANTOS

    2010-01-01

    Full Text Available Today, the Municipal Health Department is required to ensure the supply of medicines to health facilities in the city, this requires performing multiple tasks such as planning and stock control, storage and handling. Many people depend on the correct supply of medicines, so the importance of conducting a study of these activities. This work initially describes the activities currently carried out in storage and distribution of medicines. These activities were studied through site observations, measurements and interviews with workers. After collecting and analyzing data, there were problems on several factors, such as conditions of shelves and storage equipment available, flow, assembly and dispatch of orders. The result of the study is summarized in the proposition of a new layout that allows for adequate storage of Brazilian norms, suggesting new facilities and appropriate equipment and safer handling and storage, special packaging for pharmaceuticals, standardizing and adapting forms of storage at good storage practices for drugs.

  13. STUDY OF A PUBLIC STORAGE AND HANDLING SYSTEM OF MEDICINES

    OpenAIRE

    KEILA RAQUEL DOS SANTOS; ÉRICO DANIEL RICARDI GUERREIRO

    2010-01-01

    Today, the Municipal Health Department is required to ensure the supply of medicines to health facilities in the city, this requires performing multiple tasks such as planning and stock control, storage and handling. Many people depend on the correct supply of medicines, so the importance of conducting a study of these activities. This work initially describes the activities currently carried out in storage and distribution of medicines. These activities were studied through site observations...

  14. Design, production and initial state of the canister

    Energy Technology Data Exchange (ETDEWEB)

    Cederqvist, Lars; Johansson, Magnus; Leskinen, Nina; Ronneteg, Ulf

    2010-12-15

    The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility.The report provides input on the initial state of the canisters to the assessment of the long-term safety, SR-Site. The initial state refers to the properties of the engineered barriers once they have been finally placed in the KBS-3 repository and will not be further handled within the repository facility. In addition, the report provides input to the operational safety report, SR-Operation, on how the canisters shall be handled and disposed. The report presents the design premises and reference design of the canister and verifies the conformity of the reference design to the design premises. The production methods and the ability to produce canisters according to the reference design are described. Finally, the initial state of the canisters and their conformity to the reference design and design premises are presented

  15. Design, production and initial state of the canister

    International Nuclear Information System (INIS)

    The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility.The report provides input on the initial state of the canisters to the assessment of the long-term safety, SR-Site. The initial state refers to the properties of the engineered barriers once they have been finally placed in the KBS-3 repository and will not be further handled within the repository facility. In addition, the report provides input to the operational safety report, SR-Operation, on how the canisters shall be handled and disposed. The report presents the design premises and reference design of the canister and verifies the conformity of the reference design to the design premises. The production methods and the ability to produce canisters according to the reference design are described. Finally, the initial state of the canisters and their conformity to the reference design and design premises are presented

  16. Contact-handled transuranic transportation system structural analysis

    International Nuclear Information System (INIS)

    The Transuranic Package Transporter (TRUPACT) is a Type B overpack being developed for contact-handled transuranic waste. End-on, side-on, and corner impacts of the loaded TRUPACT due to a 9 m drop onto an unyielding surface have been analyzed. In each case the analyses progressed from simplified hand approaches to successively more complex finite element calculations. The first analysis of each series represents the hand calculations which were carried out to obtain initial thicknesses of foam. The remaining analyses were performed using the dynamic and nonlinear analysis capabilities of ADINA, a structural analysis finite element computer program

  17. SIMULASI GROUP TECHNOLOGY SYSTEM UNTUK MEMINIMALKAN BIAYA MATERIAL HANDLING DENGAN METODE HEURISTIC

    OpenAIRE

    Much Djunaidi; Munajat Tri Nugroho; Johan Anton

    2006-01-01

    Group Technology System merupakan metode pengaturan fasilitas produksi (machine groups) yang dibutuhkan untuk memproses suatu part family tertentu ke dalam sel-sel manufaktur. Pengaturan tata letak di CV. Sonytex yang berdasarkan process layout mengakibatkan perusahaan menghadapi permasalahan berupa tingginya kebutuhan material handling. Salah satu kriteria kinerja dalam pembentukan sel manufaktur pada GTS adalah meminimasi total jarak material handling, sehingga dapat mengurangi biaya materi...

  18. Learning to Design and Analyze Materials Handling Systems: Developing Multimedia Tools

    Science.gov (United States)

    Heragu, Sunderesh; Jennings, Sybillyn

    2003-01-01

    In this paper, we describe aspects related to learning and learning assessment including pedagogy, cognition, pilot study and results from the study. This study is conducted for an educational module on "10 Principles of Materials Handling". This module along with another on "Analysis and Design of Integrated Materials Handling Systems" constitute…

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

    OpenAIRE

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

    2013-01-01

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

  20. Stress analysis of high-level waste canisters: methods, applications, and design data

    International Nuclear Information System (INIS)

    An overview of stress analysis methods, structural design procedures, and design data is presented for canisters used to package solidified wastes, particularly borosilicate glass. In addition, waste processing, canister materials, fabrication and inspection methods, and performance testing are summarized. Sources of stress in canisters are lifting and handling loads, internal pressure, high-temperature filling operations, transient heating and cooling, differential thermal expansions of canisters and glass, and impact loadings from low-probability accidents. Results of case studies that illustrate applicable methods of stress analyses are presented for these sources of stress. Existing sections of ASME Boiler and Pressure Vessel Code are applicable to canister fabrication, but the code does not cover many aspects of canister service loadings. Specialized criteria for minimum wall thicknesses to sustain filling stresses are proposed in this report. Results of a test program to measure the creep strength of candidate canister materials are described. Methods to predict residual stresses in the walls of waste canisters are described; predicted residual stress levels agree with measured stress levels. The consequences of these residual stresses are reviewed, and stress-corrosion cracking is identified as the mode of canister failure affected by residual stresses. Canister-closure design is covered in detail, particularly the welding and inspection of the final closure seal-weld. It is shown that the methods of fracture mechanics and fatigue-crack-growth analyses are valuable tools for evaluating the performance of closure welds in the presence of crack-like defects. Canister performance in process trials at PNL shows the ability of canisters to survive high temperatures and loadings during processing. Impact tests show that a suitably designed canister can sustain severe impacts without loss of intergrity

  1. Moisture insensitive charcoal canisters

    International Nuclear Information System (INIS)

    Continuous monitoring of 222Rn concentrations in the air in houses is the most appropriate approach for the real-time measurements, but this requires complex and expensive instruments and is not practical for large studies. Activated carbon canisters have been used extensively for determining the average concentration over a period of a few days. The ''open face'' charcoal detectors have an integration time constant of about 14 h so that they are sensitive to short-term transient changes in the radon concentration. In addition, water uptake at high relative humidities reduces the radon uptake by the charcoal. The addition of a diffusion barrier and a nylon screen results in a charcoal detector with an integration half-time ranging from 20 to 60 h and a reduced uptake of water at high humidities. Silicone rubber sheeting is relatively permeable to radon and impermeable to water vapor. It was the purpose of this study to evaluate the effect of a silicone barrier on the charcoal canister radon collective device. 3 refs

  2. Event detection and exception handling strategies in the ASDEX Upgrade discharge control system

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de; Neu, G.; Rapson, C.; Raupp, G.; Zasche, D.; Zehetbauer, T.

    2013-10-15

    Highlights: •Event detection and exception handling is integrated in control system architecture. •Pulse control with local exception handling and pulse supervision with central exception handling are strictly separated. •Local exception handling limits the effect of an exception to a minimal part of the controlled system. •Central Exception Handling solves problems requiring coordinated action of multiple control components. -- Abstract: Thermonuclear plasmas are governed by nonlinear characteristics: plasma operation can be classified into scenarios with pronounced features like L and H-mode, ELMs or MHD activity. Transitions between them may be treated as events. Similarly, technical systems are also subject to events such as failure of measurement sensors, actuator saturation or violation of machine and plant operation limits. Such situations often are handled with a mixture of pulse abortion and iteratively improved pulse schedule reference programming. In case of protection-relevant events, however, the complexity of even a medium-sized device as ASDEX Upgrade requires a sophisticated and coordinated shutdown procedure rather than a simple stop of the pulse. The detection of events and their intelligent handling by the control system has been shown to be valuable also in terms of saving experiment time and cost. This paper outlines how ASDEX Upgrade's discharge control system (DCS) detects events and handles exceptions in two stages: locally and centrally. The goal of local exception handling is to limit the effect of an unexpected or asynchronous event to a minimal part of the controlled system. Thus, local exception handling facilitates robustness to failures but keeps the decision structures lean. A central state machine deals with exceptions requiring coordinated action of multiple control components. DCS implements the state machine by means of pulse schedule segments containing pre-programmed waveforms to define discharge goal and control

  3. Event detection and exception handling strategies in the ASDEX Upgrade discharge control system

    International Nuclear Information System (INIS)

    Highlights: •Event detection and exception handling is integrated in control system architecture. •Pulse control with local exception handling and pulse supervision with central exception handling are strictly separated. •Local exception handling limits the effect of an exception to a minimal part of the controlled system. •Central Exception Handling solves problems requiring coordinated action of multiple control components. -- Abstract: Thermonuclear plasmas are governed by nonlinear characteristics: plasma operation can be classified into scenarios with pronounced features like L and H-mode, ELMs or MHD activity. Transitions between them may be treated as events. Similarly, technical systems are also subject to events such as failure of measurement sensors, actuator saturation or violation of machine and plant operation limits. Such situations often are handled with a mixture of pulse abortion and iteratively improved pulse schedule reference programming. In case of protection-relevant events, however, the complexity of even a medium-sized device as ASDEX Upgrade requires a sophisticated and coordinated shutdown procedure rather than a simple stop of the pulse. The detection of events and their intelligent handling by the control system has been shown to be valuable also in terms of saving experiment time and cost. This paper outlines how ASDEX Upgrade's discharge control system (DCS) detects events and handles exceptions in two stages: locally and centrally. The goal of local exception handling is to limit the effect of an unexpected or asynchronous event to a minimal part of the controlled system. Thus, local exception handling facilitates robustness to failures but keeps the decision structures lean. A central state machine deals with exceptions requiring coordinated action of multiple control components. DCS implements the state machine by means of pulse schedule segments containing pre-programmed waveforms to define discharge goal and control

  4. Material handling systems for the fluidized-bed combustion boiler at Rivesville, West Virginia

    Science.gov (United States)

    Branam, J. G.; Rosborough, W. W.

    1977-01-01

    The 300,000 lbs/hr steam capacity multicell fluidized-bed boiler (MFB) utilizes complex material handling systems. The material handling systems can be divided into the following areas: (1) coal preparation; transfer and delivery, (2) limestone handling system, (3) fly-ash removal and (4) bed material handling system. Each of the above systems are described in detail and some of the potential problem areas are discussed. A major potential problem that exists is the coal drying system. The coal dryer is designed to use 600 F preheated combustion air as drying medium and the dryer effluent is designed to enter a hot electrostatic precipitator (730 F) after passage through a cyclone. Other problem areas to be discussed include the steam generator coal and limestone feed system which may have operating difficulties with wet coal and/or coal fines.

  5. Fuel Handling Systems: Some technical orientations for ASTRID project

    International Nuclear Information System (INIS)

    This article deals with the technical orientations considered for fuel handling management in ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project. Its purpose is to present the orientations proposed by AREVA to solve the technical and economical issues involved by sodium fast neutrons reactors and by ASTRID specificities. In the reactor vessel, the fuel handling is carried out under rotating plugs by means of a direct lift charge machine and a fixed arm charge machine. To unload a minor actinide assembly as quick as required by ASTRID objectives, a sodium way has been chosen for core loading - unloading. A spent fuel/new fuel exchange device is implemented inside the reactor vessel to increase core loading - unloading efficiency and to improve the global plant availability. The technical feasibility of an external storage in sodium is discussed and a first design proposed. Management of new, spent and failed fuel from external storage to the outlet of the nuclear island is also described and a general architecture is proposed. (author)

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

    International Nuclear Information System (INIS)

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

  7. SIMULASI GROUP TECHNOLOGY SYSTEM UNTUK MEMINIMALKAN BIAYA MATERIAL HANDLING DENGAN METODE HEURISTIC

    Directory of Open Access Journals (Sweden)

    Much. Djunaidi

    2006-04-01

    Full Text Available Group Technology System merupakan metode pengaturan fasilitas produksi (machine groups yang dibutuhkan untuk memproses suatu part family tertentu ke dalam sel-sel manufaktur. Pengaturan tata letak di CV. Sonytex yang berdasarkan process layout mengakibatkan perusahaan menghadapi permasalahan berupa tingginya kebutuhan material handling. Salah satu kriteria kinerja dalam pembentukan sel manufaktur pada GTS adalah meminimasi total jarak material handling, sehingga dapat mengurangi biaya material handling dan meningkatkan produktivitas. Dalam penelitian ini digunakan tiga metode, yaitu Bond Energy Algorithm (BEA, Rank Order Clustering (ROC dan Rank Order Clustering 2 (ROC2. Hasil dari penelitian ini adalah dengan menerapkan group technology systems diperoleh total pengurangan jarak material handling sebesar 70 m dan penghematan biaya material handling sebesar Rp 1.534.978,-. Berdasarkan model simulasi, relayout dengan metode BEA meningkatkan jumlah produksi sebesar 1 unit produk/hari dan penurunan waktu tunggu sebesar 0,575 menit.

  8. Planning and control of automated material handling systems: The merge module

    OpenAIRE

    Haneyah, Sameh; Hurink, Johann; Schutten, Marco; Zijm, Henk; Schuur, Peter; Hu, Bo; Morasch, Karl; Stefan PICKL; Siegle, Markus

    2011-01-01

    We address the field of internal logistics, embodied in Automated Material Handling Systems (AMHSs), which are complex installations employed in sectors such as Baggage Handling, Physical Distribution, and Parcel & Postal. We work on designing an integral planning and real-time control architecture, and a set of generic algorithms for AMHSs. Planning and control of these systems need to be robust, and to yield close-to-optimal system performance. Currently, planning and control of AMHSs is hi...

  9. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

    Energy Technology Data Exchange (ETDEWEB)

    HOLLENBECK, R.G.

    2000-05-08

    The Spent Nuclear Fuel (SNF) Canister Storage Building (CSB) is the interim storage facility for the K-Basin SNF at the US. Department of Energy (DOE) Hanford Site. The SNF is packaged in multi-canister overpacks (MCOs). The MCOs are placed inside transport casks, then delivered to the service station inside the CSB. At the service station, the MCO handling machine (MHM) moves the MCO from the cask to a storage tube or one of two sample/weld stations. There are 220 standard storage tubes and six overpack storage tubes in a below grade reinforced concrete vault. Each storage tube can hold two MCOs.

  10. Techniques for freeing deposited canisters. Final report

    International Nuclear Information System (INIS)

    frequency alternating current technique. Little information was found, however, regarding the interaction between high frequency alternating current and bentonite. It could nonetheless be assessed that the technique would be associated with a high degree of complexity as well as a high demand for energy/power. None of the methods studied for determining the position of the canister could be assessed to have any significant potential for an accurate determination of the position of the canister in an opened deposition hole. The conclusion is that the choice of methods for freeing should be focussed on methods which do not require any detailed determination of the position of the canister. A number of generic criteria were identified and used in the subsequent categorization of the different techniques for freeing. After the evaluation the techniques were divided into three groups: Techniques which have a high potential for development of a system for freeing of the canister. Techniques which have a low potential for development of a system for freeing of the canister. Techniques which are not recommended for further investigation. Only one technique was identified in the high potential category, namely the low-pressure hydrodynamic technique. Four techniques were identified to have a low potential (cooling of the buffer, cooling of the canister, water jet technique and application of direct current). The other seven techniques included are not recommended for further studies. Since the comparison had to be based on a simple and generic set of criteria, a further analysis was made in order to determine whether or not the low pressure hydrodynamic method is robust enough in order to remain in the high potential category even when some other relevant issues are considered. This was found to be the case

  11. Summary of Preliminary Criticality Analysis for Peach Bottom Fuel in the DOE Standardized Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    The Department of Energy's (DOE's) National Spent Nuclear Fuel Program is developing a standardized set of canisters for DOE spent nuclear fuel (SNF). These canisters will be used for DOE SNF handling, interim storage, transportation, and disposal in the national repository. Several fuels are being examined in conjunction with the DOE SNF canisters. This report summarizes the preliminary criticality safety analysis that addresses general fissile loading limits for Peach Bottom graphite fuel in the DOE SNF canister. The canister is considered both alone and inside the 5-HLW/DOE Long Spent Fuel Co-disposal Waste Package, and in intact and degraded conditions. Results are appropriate for a single DOE SNF canister. Specific facilities, equipment, canister internal structures, and scenarios for handling, storage, and transportation have not yet been defined and are not evaluated in this analysis. The analysis assumes that the DOE SNF canister is designed so that it maintains reasonable geometric integrity. Parameters important to the results are the canister outer diameter, inner diameter, and wall thickness. These parameters are assumed to have nominal dimensions of 45.7-cm (18.0-in.), 43.815-cm (17.25-in), and 0.953-cm (0.375-in.), respectively. Based on the analysis results, the recommended fissile loading for the DOE SNF canister is 13 Peach Bottom fuel elements if no internal steel is present, and 15 Peach Bottom fuel elements if credit is taken for internal steel

  12. 78 FR 41810 - Proposed Revisions to Light Load Handling System and Operations

    Science.gov (United States)

    2013-07-11

    ... operating experience associated with Bulletin 84-03, ``Refueling Cavity Water Seals'' (ADAMS Accession No... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Proposed Revisions to Light Load Handling System and Operations AGENCY: Nuclear...

  13. Sample Handling System for in-situ Powder X-ray Diffraction Instruments. Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed innovation is a Powder Handling System (PHS) that will deliver powdered samples to in situ planetary XRD instruments and provide unique means of...

  14. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    The holding canister for spent nuclear fuel will be transferred by a lift to the final disposal tunnels 500m deep in the bedrock. Model tests were carried out with an objective to estimate weather feasible shock absorbing properties can be met in a design accident case where the canister should survive a free fall due to e.g. sabotage. If the velocity of the canister is not controlled by air drag or any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity when stricken by the surface penetration impact if the bottom pit of the lift well would be filled with groundwater. However the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20m high filling to the bottom pit of the lift well by ceramic gravel, trade mark LECA-sora, gives a fair impact absorption to protect the spent fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)

  15. Defense High Level Waste Disposal Container System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    N. E. Pettit

    2001-07-13

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by

  16. Selection and Implementation of an Optimal System to Handle Garbage in Kigali, Rwanda

    OpenAIRE

    Innocent, Kahigana

    2011-01-01

    Reports from various institutions claim that garbage management in Rwanda has had diverse effects on both the natural environment and human society. Such claims prompted for an exploratory study to find out an optimal system to handle solid waste in Kigali City. The study considered a literature review and primary data from 400 randomly selected citizens. They were surveyed about their opinions on which system they perceived to be the optimal to handle garbage in Kigali City. The computer sof...

  17. Ventilation system of actinides handling facility in Oarai-branch of Tohoku University

    International Nuclear Information System (INIS)

    We have reported the development of the facility for handling actinides in Tohoku University at the second KAERI-JAERI joint seminar on PIE technology. Actinide isotopes have most hazardous α-radioactivity. Therefore, a specially designed facility is necessary to carry out experimental study for actinide physics and chemistry. In this paper, we will describe the ventilation system and monitoring system for actinide handling facility. (author)

  18. Specification of failure-handling requirements as policy rules on self-adaptive systems

    OpenAIRE

    Pimentel, João Henrique; Castro, Jaelson; Franch Gutiérrez, Javier

    2012-01-01

    Most adaptive systems have compensation mechanisms for recovering from or preventing failures. However, sometimes a compensation is not essential. Hence, diagnosing and compensating each and every one of their failures may be ineffective. Rather than polluting a requirements specification with fine grained definition of failure-handling conditions, this work aims to increase the flexibility of failure handling in self-adaptive systems using tolerance policies. We allow the expression...

  19. Comments on 'SKB FUD-program 95' focused on canister integrity and corrosion

    International Nuclear Information System (INIS)

    The work presented in this report is a result of reading the SKB program for R,D and D on safe storage of radioactive wastes. Our work, which is focused on the waste canisters, was commissioned by the Swedish Nuclear Power Inspectorate. We find the program very difficult to follow owing to the lack of detail in chapter seven. In our opinion this will make the work difficult to monitor by SKI or SKB. We also feel that the interpretation of information already available is overoptimistic. As a consequence the difficulties ahead are understated and the programme is converging too quickly. We believe that it should be possible to develop a satisfactory canister for disposal of high level nuclear waste according to the general method proposed by SKB and with the proposed capacity within the timescale of the overall programme. We do not believe, however, that all the difficulties have been recognised. As a consequence of this the results to date are interpreted optimistically. We believe that progress should be subjected to more professional review within SKB and that a higher level of metallurgical support is required. We disagree that suitable full size canisters have been created and that production technology is available for both canisters at full size. We also disagree that the long-time durability is ascertained. I.a. it is easy to find corrosion mechanisms for the canister system that have to be demonstrated not to be harmful. We feel there are many areas which need further evaluation, i.a. effects of non uniform loading and creep, effects of departure from circularity, welding, quality control, effects of radiolysis, corrosion properties, etc. We also feel that insufficient emphasis has been placed on the further development on high power electron beam welding, machining, casting of the insert, testing and overall handling. We consider that more information should be provided on the detail and timing of the development plan for the trial fabrication programme of

  20. Reliability in sealing of canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    obtained with NDT. The predicted maximum discontinuity size in connection with the welding of 4,500 canisters at the present stage of development of the process was conservatively determined to be less than one centimetre. All factors considered, the predicted minimum copper coverage for a 5 cm thick canister is 4 cm. Acceptance criteria for permitted settings in the welding process in a future sealing system are proposed, as is the use of statistical process control based on nondestructive testing as an independent inspection system. Furthermore, principles for handling of process non conformances are presented

  1. Reliability in sealing of canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ronneteg, Ulf [Bodycote Materials Testing AB, Nykoeping (Sweden); Cederqvist, Lars; Ryden, Haakan [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Oeberg, Tomas [Tomas Oeberg Konsult AB, Karlskrona (Sweden); Mueller, Christina [Federal Inst. for Materials Research and Testing, Berlin (Germany)

    2006-06-15

    obtained with NDT. The predicted maximum discontinuity size in connection with the welding of 4,500 canisters at the present stage of development of the process was conservatively determined to be less than one centimetre. All factors considered, the predicted minimum copper coverage for a 5 cm thick canister is 4 cm. Acceptance criteria for permitted settings in the welding process in a future sealing system are proposed, as is the use of statistical process control based on nondestructive testing as an independent inspection system. Furthermore, principles for handling of process non conformances are presented.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Safety assessment of a robotic system handling nuclear material

    International Nuclear Information System (INIS)

    This paper outlines the use of a Failure Modes and Effects Analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, The Weigh and Leak Check System, is to replace a manual process at the Department of Energy facility at Pantex by which nuclear material is inspected for weight and leakage. Failure Modes and Effects Analyses were completed for the robotics process to ensure that safety goals for the system had been meet. These analyses showed that the risks to people and the internal and external environment were acceptable

  6. Inspection of copper canisters for spent nuclear fuel by means of ultrasonic array system. Modelling, defect detection and grain noise estimation

    International Nuclear Information System (INIS)

    The work presented in the report has been split into three overlapping tasks which have the following objectives: (1) development of beam-forming tools, and verification of modeling tools; (2) investigation of detection and resolution limits; (3) evaluation of attenuation, estimation and suppression of grain noise. For beam-forming tools, a method of designing steered and/or focused beams in immersed solids is presented based on geometrical acoustics. Presently, the beam designs are only related to delays but not to apodization. These focused, steered beams are intended to be used for sizing defects and inspecting the regions close to canisters outer walls. The modeling tool developed previously for simulating elastic fields radiated by planar arrays into immersed solids has been verified by comparing with the results obtained from PASS, a software developed by Dr. Didier Cassereau, France. The results from our modeling tool are in excellent agreement with those from PASS. Since the array coming with the ALLIN ultrasonic array system is not planar, but cylindrically curved in elevation, and it works not in transmission mode, but in pulse echo mode, the above modeling tool for the planar arrays cannot be applied directly. Therefore, the modeling tool has been upgraded for the ALLIN array. The theory underlying this modeling tool is the extended angular spectrum approach (ASA) which was developed based on the conventional ASA that only applies to planar sources. Experimental verification of the modeling tool has shown that the results from the tool agree very well with the measurements. To quantify the fields from the ALLIN array and to facilitate the comparison of simulated results with the measured ones, the ALLIN array system has been calibrated based on the existing functionality, and an analytical model has been proposed for simulating measured acoustic echo pulses. To investigate the detection and resolution limits, we have carried out a series of experiments

  7. Inspection of copper canisters for spent nuclear fuel by means of ultrasonic array system. Modelling, defect detection and grain noise estimation

    Energy Technology Data Exchange (ETDEWEB)

    Wu Ping; Stepinski, T. [Uppsala Univ., (Sweden). Dept. of Material Science

    1998-07-01

    The work presented in the report has been split into three overlapping tasks which have the following objectives: (1) development of beam-forming tools, and verification of modeling tools; (2) investigation of detection and resolution limits; (3) evaluation of attenuation, estimation and suppression of grain noise. For beam-forming tools, a method of designing steered and/or focused beams in immersed solids is presented based on geometrical acoustics. Presently, the beam designs are only related to delays but not to apodization. These focused, steered beams are intended to be used for sizing defects and inspecting the regions close to canisters outer walls. The modeling tool developed previously for simulating elastic fields radiated by planar arrays into immersed solids has been verified by comparing with the results obtained from PASS, a software developed by Dr. Didier Cassereau, France. The results from our modeling tool are in excellent agreement with those from PASS. Since the array coming with the ALLIN ultrasonic array system is not planar, but cylindrically curved in elevation, and it works not in transmission mode, but in pulse echo mode, the above modeling tool for the planar arrays cannot be applied directly. Therefore, the modeling tool has been upgraded for the ALLIN array. The theory underlying this modeling tool is the extended angular spectrum approach (ASA) which was developed based on the conventional ASA that only applies to planar sources. Experimental verification of the modeling tool has shown that the results from the tool agree very well with the measurements. To quantify the fields from the ALLIN array and to facilitate the comparison of simulated results with the measured ones, the ALLIN array system has been calibrated based on the existing functionality, and an analytical model has been proposed for simulating measured acoustic echo pulses. To investigate the detection and resolution limits, we have carried out a series of experiments

  8. Data handling in safety monitoring system of nuclear power plant

    International Nuclear Information System (INIS)

    The method about the real-time dynamic system parameter's acquisition, unit change, scale change, data reduction of communication and special parameter calculation are described. The programming structure chart about the data processing is given. The importance of data processing in dynamic systems is pointed

  9. Communication-based fault handling scheme for ungrounded distribution systems

    International Nuclear Information System (INIS)

    The requirement for high quality and highly reliable power supplies has been increasing as a result of increasing demand for power. At the time of a fault occurrence in a distribution system, some protection method would be dedicated to fault section isolation and service restoration. However, if there are many outage areas when the protection method is performed, it is an inconvenience to the customer. A conventional method to determine a fault section in ungrounded systems requires many successive outage invocations. This paper proposed an efficient fault section isolation method and service restoration method for single line-to-ground fault in an ungrounded distribution system that was faster than the conventional one using the information exchange between connected feeders. The proposed algorithm could be performed without any power supply interruption and could decrease the number of switching operations, so that customers would not experience outages very frequently. The method involved the use of an intelligent communication method and a sequential switching control scheme. The proposed algorithm was also applied in both a single-tie and multi-tie distribution system. This proposed algorithm has been verified through fault simulations in a simple model of ungrounded multi-tie distribution system. The method proposed in this paper was proven to offer more efficient fault identification and much less outage time than the conventional method. The proposed method could contribute to a system design since it is valid in multi-tie systems. 5 refs., 2 tabs., 8 figs

  10. Fluid (Air/Water) Cushion Transportation Technology for Emplacing Heavy Canisters into Horizontal Disposal Drifts

    International Nuclear Information System (INIS)

    The disposal of certain types of radioactive waste canisters in a deep repository involves handling and emplacement of very heavy loads. The weight of these particular canisters can be in the order of 20 to 50 metric tons. They generally have to be handled underground in openings that are not much larger than the canisters themselves as it is time consuming and expensive to excavate and backfill large openings in a repository. This therefore calls for the development of special technology that can meet the requirements for safe operation at an industrial scale in restrained operating spaces. Air/water cushion lifting systems are used world wide in the industry for moving heavy loads. However, until now the technology needed for emplacing heavy cylindrical radioactive waste packages in bored drifts (with narrow annular gaps) has not been previously developed or demonstrated. This paper describes the related R and D work carried out by ANDRA (for air cushion technology) and by SKB and Posiva (for water cushion technology) respectively, mainly within the framework of the European Commission (EC) funded Integrated Project called ESDRED (6. European Framework Programme). The background for both the air and the water cushion applications is presented. The specific characteristics of the two different emplacement concepts are also elaborated. Then the various phases of the Test Programmes (including the Prototype phases) are detailed and illustrated for the two lifting media. Conclusions are drawn for each system developed and evaluated. Finally, based on the R and D experience, improvements deemed necessary for an industrial application are listed. The tests performed so far have shown that the emplacement equipment developed is operating efficiently. However further tests are required to verify the availability and the reliability of the equipment over longer periods of time and to identify the modifications that would be needed for an industrial application in a

  11. Bifurcation methods of dynamical systems for handling nonlinear wave equations

    Indian Academy of Sciences (India)

    Dahe Feng; Jibin Li

    2007-05-01

    By using the bifurcation theory and methods of dynamical systems to construct the exact travelling wave solutions for nonlinear wave equations, some new soliton solutions, kink (anti-kink) solutions and periodic solutions with double period are obtained.

  12. Using Self-Description to Handle Change in Systems

    CERN Document Server

    Estrella, Florida; Le Goff, Jean-Marie; McClatchey, Richard; Murray, Steven

    2002-01-01

    In the web age systems must be flexible, reconfigurable and adaptable in addition to being quick to develop. As a consequence, designing systems to cater for change is becoming not only desirable but required by industry. Allowing systems to be self-describing or description-driven is one way to enable these characteristics. To address the issue of evolvability in designing self-describing systems, this paper proposes a pattern-based, object-oriented, description-driven architecture. The proposed architecture embodies four pillars - first, the adoption of a multi-layered meta-modeling architecture and reflective meta-level architecture, second, the identification of four data modeling relationships that must be made explicit such that they can be examined and modified dynamically, third, the identification of five design patterns which have emerged from practice and have proved essential in providing reusable building blocks for data management, and fourth, the encoding of the structural properties of the fiv...

  13. The beam handling system of the Oslo Cyclotron

    International Nuclear Information System (INIS)

    The beam optic system of the Oslo Cyclotron is described. A computer program for the calculation of optimal settings of quadropoles is presented. The reliability of the computer program is confirmed by experimental data

  14. Problem Handling and Improvement of Imported SCR Static Excitation System

    Institute of Scientific and Technical Information of China (English)

    Hao Shanwu; Li Weimin; Liang Jianfen; Chu Xue

    2006-01-01

    @@ Static self-excitation system has been more and more widely used for large turbogenerator units in China and achieved very good results in recent years. However, new problems have arisen along with the development of high-power thyristor technology, etc. In view of the actual operating condition of SEE900/5000 SCR static excitation system imported from Siemens, Germany, some technical renovations were carried out.

  15. Replacement of heavy water supply controllers in fuel handling system

    International Nuclear Information System (INIS)

    Full text: Both automatic analog controllers and both analog manual controllers (63526-PC no. 11 A and 63526-PC11 no. 1 C) (63526-PC11 no. 2 A and 63526-PC11 no. 2 C) were replaced by three Digital Controllers as follow: 63526-PC11A, 63526-PC11C and 63526-PC11 because of the aging of old analog controller. In the new System, there is a separate controllers for each one in both side, (Side A and Side C) and other to control the common bleed valve. The reasons more important for this replace. The analog pressure control system: - Is a design from the early 1980s; - Analog control principle used; - Custom-built instrument control system; - System control by electronic hardware. No programming involved; - There is not more commercial support for hardware. (electronic printed circuit board). The digital pressure control system: - Programmable; - Commercial full support; - Easy calibration and maintenance; - Easy operation. Because of the above keys, Embalse NPP decided to change the control system together with AECL, without shutdown the Plant. The last one was the challenge, to install and commissioning the new controllers without affect the electrical production. (author)

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

    International Nuclear Information System (INIS)

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

  17. A Morphological Box for Handling Temporal Data in B2C Systems

    OpenAIRE

    Knolmayer, Gerhard F.; Borean, Alessandro

    2010-01-01

    User interfaces are key properties of Business-to-Consumer (B2C) systems, and Web-based reservation systems are an important class of B2C systems. In this paper we show that these systems use a surprisingly broad spectrum of different approaches to handling temporal data in their Web inter faces. Based on these observations and on a literature analysis we develop a Morphological Box to present the main options for handling temporal data and give examples. The results indicate that the present...

  18. The design of in-cell crane handling systems for nuclear plants

    International Nuclear Information System (INIS)

    The reprocessing and waste management facilities at (BNFL's) British Nuclear Fuels Limited's Sellafield site make extensive use of crane handling systems. These range from conventional mechanical handling operations as used generally in industry to high integrity applications through to remote robotic handling operations in radiation environments. This paper describes the design methodologies developed for the design of crane systems for remote handling operations - in-cell crane systems. In most applications the in-cell crane systems are an integral part of the plant process equipment and reliable and safe operations are a key design parameter. Outlined are the techniques developed to achieve high levels of crane system availability for operations in hazardous radiation environments. These techniques are now well established and proven through many years of successful plant operation. A recent application of in-cell crane handling systems design for process duty application is described. The benefits of a systematic design approach and a functionally-based engineering organization are also highlighted. (author)

  19. Online Decision Support System (IRODOS) - an emergency preparedness tool for handling offsite nuclear emergency

    International Nuclear Information System (INIS)

    A real time online decision support system as a nuclear emergency response system for handling offsite nuclear emergency at the Nuclear Power Plants (NPPs) has been developed by Health, Safety and Environment Group, Bhabha Atomic Research Centre (BARC), Department of Atomic Energy (DAE) under the frame work of 'Indian Real time Online Decision Support System 'IRODOS'. (author)

  20. Thermal Dimensioning of SiC Canister Applied A-KRS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Inyoung; Choi, Heuijoo; Yoo, Malgobalgebitnala [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Reducing toxicity and volume of SNF and reusing valuable fissile materials, pyro-processing connected with SFR is under-developing. The A-KRS is composed of 1 cm thick copper cold-spray-coated cast iron canisters, buffer blocks, disposal holes and disposal tunnels, etc. To manufacture disposal canisters, massive un-reusable copper and iron resources are required. Recently, SiC which has high thermal conductivity and good mechanical properties is investigated as a substitute material of metal canister to save metal resources. In this study, thermal performance of SiC canister is investigated and thermal dimensioning of SiC canister applied A-KRS is conducted to estimate thermal applicability of SiC canister in repository. In this study, thermal applicability of SiC as a substitute material of copper and cast iron canister is assessed. Due to higher thermal conductivity of SiC, calculated maximum temperature of SiC applied system is lower than original metal canister applied system and estimated minimum disposal hole pitch of SiC canister system is narrower than metal canister system. But decrease of distance between disposal hole pitch by adopting SiC canister is negligible considering engineering and safety margin. As a result, it is confirmed that SiC could be used as a substitute materials of metal in respect of thermal aspect. To apply SiC canister in deep geological repository, however, thermal-mechanical assessment need to be conducted as future studies. Especially thermally induced stress and intactness of canister must be estimated because SiC is fragile material and its thermal conductivity is highly dependent on temperature.

  1. Thermal Dimensioning of SiC Canister Applied A-KRS

    International Nuclear Information System (INIS)

    Reducing toxicity and volume of SNF and reusing valuable fissile materials, pyro-processing connected with SFR is under-developing. The A-KRS is composed of 1 cm thick copper cold-spray-coated cast iron canisters, buffer blocks, disposal holes and disposal tunnels, etc. To manufacture disposal canisters, massive un-reusable copper and iron resources are required. Recently, SiC which has high thermal conductivity and good mechanical properties is investigated as a substitute material of metal canister to save metal resources. In this study, thermal performance of SiC canister is investigated and thermal dimensioning of SiC canister applied A-KRS is conducted to estimate thermal applicability of SiC canister in repository. In this study, thermal applicability of SiC as a substitute material of copper and cast iron canister is assessed. Due to higher thermal conductivity of SiC, calculated maximum temperature of SiC applied system is lower than original metal canister applied system and estimated minimum disposal hole pitch of SiC canister system is narrower than metal canister system. But decrease of distance between disposal hole pitch by adopting SiC canister is negligible considering engineering and safety margin. As a result, it is confirmed that SiC could be used as a substitute materials of metal in respect of thermal aspect. To apply SiC canister in deep geological repository, however, thermal-mechanical assessment need to be conducted as future studies. Especially thermally induced stress and intactness of canister must be estimated because SiC is fragile material and its thermal conductivity is highly dependent on temperature

  2. Computer systems handle changing T-line access

    Energy Technology Data Exchange (ETDEWEB)

    Paula, G.

    1990-03-01

    As the number of power exchanges among utilities grows and transmission lines are loaded more heavily, it becomes increasingly difficult to manage power-system access. A study sponsored by the Electric Power Research Institute (EPRI) identifies two analysis techniques that can provide more detailed line-use information to help utilities ensure continued reliability. After meeting internal needs, a utility agrees on usage price and terms with other power suppliers and users that want to transfer power across its lines. This is known as wheeling. However, such transactions affect the loading of lines belonging to other utilities. As a result, no utility can actually control who uses its transmission system. Many utilities would like a way to monitor power flows on their systems to improve the economy and reliability of operation. The EPRI-sponsored study, conducted by Casazza, Schultz, Associates (CSA), Arlington, VA, identifies ways that computer methods can help utilities cope with increased line access.

  3. A distributed microprocessor system for spacecraft control and data handling

    Science.gov (United States)

    Rennels, D. A.

    1977-01-01

    The specific requirements for spacecraft computing systems are considered. These requirements are partly related to the constraints of limited resources of power, weight, and volume. Another important factor is the requirement of extremely high reliability. These reliability requirements have led to introduction of automated redundancy techniques on board the spacecraft. The various redundant computers check each other and provide recovery procedures when a computer is found to have failed. Past and future capabilities are considered along with distributed processing requirements. System considerations are discussed, taking into account suboptimum computer throughput, sensitivity to software modifications, hierarchic timing, I/O granularity, restricted communications, synchronous functions, hierarchic control, and concurrent error detection. A description is presented of the Unified Data System (UDS), which consists of a set of standard microcomputers connected by several buses. Attention is also given to synchronization and timing, the executive control structure, the programming language, and the executive program.

  4. Reliability requirements management for ITER Remote Handling maintenance systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  5. Reliability requirements management for ITER Remote Handling maintenance systems

    International Nuclear Information System (INIS)

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

  6. Handling system for nuclear reactor fuel and reflector elements

    International Nuclear Information System (INIS)

    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)

  7. A Gas Target with a Tritium Gas Handling System

    International Nuclear Information System (INIS)

    A detailed description is given of a simple tritium gas target and its tritium gas filling system, and how to put it into operation. By using the T (p,n) He reaction the gas target has been employed for production of monoenergetic fast neutrons of well defined energy and high intensity. The target has been operated successfully for a long time

  8. Planning and Control Concepts for Material Handling Systems

    NARCIS (Netherlands)

    I.F.A. Vis (Iris)

    2002-01-01

    textabstractIris Vis was born in 1974 in Leidschendam. May 2002 - Assistant professor at the School of Economics and Business Administration, Vrije Universiteit Amsterdam 1999, Visiting scholar at Georgia Institute of Technology, School of Industrial and Systems Engineering, october - december Sept

  9. Deep geological disposal system development; mechanical structural stability analysis of spent nuclear fuel disposal canister under the internal/external pressure variation

    Energy Technology Data Exchange (ETDEWEB)

    Kwen, Y. J.; Kang, S. W.; Ha, Z. Y. [Hongik University, Seoul (Korea)

    2001-04-01

    This work constitutes a summary of the research and development work made for the design and dimensioning of the canister for nuclear fuel disposal. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for spent fuel disposal should be securred. Usually this repository is expected to locate at a depth of 500m underground. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for spent nuclear fuel disposal in a deep repository in the crystalline bedrock, which entails an evenly distributed load of hydrostatic pressure from undergroundwater and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables array type of inner baskets and thicknesses of outer shell and lid and bottom are tried to be determined through the mechanical linear structural analysis, thicknesses of outer shell is determined through the nonlinear structural analysis, and the bentonite buffer analysis for the rock movement is conducted through the of nonlinear structural analysis Also the thermal stress effect is computed for the cast iron insert. The canister types studied here are one for PWR fuel and another for CANDU fuel. 23 refs., 60 figs., 23 tabs. (Author)

  10. Remote handling concept for the neutral beam system

    International Nuclear Information System (INIS)

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

  11. Applying the Mahalanobis–Taguchi System to Vehicle Handling

    OpenAIRE

    Cudney, Elizabeth A.; Paryani, Kioumars; Ragsdell, Kenneth M.

    2006-01-01

    Abstract The Mahalanobis?Taguchi system (MTS) is a diagnosis and forecasting method using multivariate data. Mahalanobis distance (MD) is a measure based on correlations between the variables and patterns that can be identified and analyzed with respect to a base or reference group. The MTS is of interest because of its reported accuracy in forecasting using small, correlated data sets. This is the type ...

  12. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    This paper covers the design, development and testing of the magazines (cylinders containing cans of plutonium-ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed robotic arm that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102mm) diameter canister throat

  13. Preliminary design specification for Department of Energy standardized spent nuclear fuel canisters. Volume 2: Rationale document

    International Nuclear Information System (INIS)

    This document (Volume 2) is a companion document to a preliminary design specification for the design of canisters to be used during the handling, storage, transportation, and repository disposal of Department of Energy (DOE) spent nuclear fuel (SNF). This document contains no procurement information, such as the number of canisters to be fabricated, explicit timeframes for deliverables, etc. However, this rationale document does provide background information and design philosophy in order to help engineers better understand the established design criteria (contained in Volume 1 respectively) necessary to correctly design and fabricate these DOE SNF canisters

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

    International Nuclear Information System (INIS)

    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

  15. Operating experiences in fuel handling system at KGS

    International Nuclear Information System (INIS)

    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)

  16. The cryogenic target handling system for the OMEGA laser

    International Nuclear Information System (INIS)

    The next series of inertial fusion experiments will approach ignition conditions. To achieve sufficient power density to approach ignition conditions with reasonable laser power, these experiments call for cryogenic targets with a uniform condensed fuel layer and a smooth inner surface inside a thin spherical shell. To field such targets, General Atomics is designing and building the OMEGA Cryogenic Target System (OCTS) for the upgraded OMEGA laser at the University of Rochester's Laboratory for Laser Energetics (LLE). The OCTS contains subsystems that operate on targets to fill, freeze, transport, layer, characterize, position to laser confluence and expose them for laser illumination milliseconds before short time. The OCTS will fill targets with deuterium-tritium (DT) to densities of up to 0.031 mol/cm3. The pressure cell and rack that contains the filled targets is cooled to cryogenic temperatures in the fill system cryostat. The rack of targets is transferred cold by the cold transfer cryostat to the transfer station cryostat. The transfer station separates individual targets from the rack and inserts them into the moving cryostat. The moving cryostat transports a target out of the LLE tritium laboratory, layers the target and inserts the target into the center of the OMEGA target tank. Targets are characterized optically during a pause in the insertion. At shot time, the end of the cryostat is rapidly removed by high acceleration motors mounted in a port on the target tank directly opposite that used to insert the target. Prototypes of the target filling and cold transfer equipment have been built and operated with deuterium. Mounted targets were successfully field to densities of 0.026 mol/cm3. Cold transfer of high density targets into and out of the fill system with the cold transfer cryostat was also successfully carried out. (author)

  17. Handling manual of computer aided tracing system 'CATS' (version I)

    International Nuclear Information System (INIS)

    As an application of computer graphics, we have developed a code named as ''CATS'' which stands for computer aided tracing system. With CATS, many kinds of graphs and tables can be fed into the host computer as graphic data by a digitizing tablet. After editing figures on the graphic display terminal, one can obtain color pictures by hard copy unit or fair monocromatic copies by a laser printer through host computer. We employ an interactive graphical input in Japanese, so that users of ''CATS'' can easily edit figures without any knowledge on the complicated FORTRAN utility package for graphic display. The usage of ''CATS'' is summarized in this report. (author)

  18. Recuperative thermal recombining system for handling loss of coolant

    International Nuclear Information System (INIS)

    A recycle loop method and recuperative heating system is provided for thermally recombining hydrogen and oxygen, such as may be desired in connection with a dissociated gas stream from a containment vessel for a nuclear reactor under loss of coolant accident conditions. The dissociated charge stream is preferably heated by indirect heat exchange with a resulting water vapor containing combined stream to effect heat conservation and, also preferably, the recuperative heat exchange operation and the high temperature recombining reaction are carried out in a unitary zone incorporating electrical, non-flame, heat input means. A particular feature of the present system resides in having the recycle of a portion of the resulting combined stream from the recombiner section pass into admixture with the hydrogen-oxygen containing charge stream to preclude an explosive condition and further having the quantity of such recycle regulated responsive to a temperature differential being measured across the recombiner zone as an indication of the oxygen level in such charge stream

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

  1. ABSTRACTION FOCUSED SYSTEM FOR USER FRIENDLY INFORMATION HANDLING OVER WWW

    Directory of Open Access Journals (Sweden)

    Dr. Pushpa R. Suri

    2011-07-01

    Full Text Available The World Wide Web has become the medium of preference for the circulation of information by common man,teams, organizations, and social communities. Information computing is the fundamental mean by which web information is retrieved and distributed. Conventional information computing approaches continues to be the most common to search documents of potential relevance. But unfortunately these offer only an imperfect solution as many relevant documents may be missed in the crude search process. The search process is sharplyquery specific and the results blindly follow the terms entered. The proposed Abstraction Focused framework for improved information computing over web attempts to resolve this basic problem that stamps from the information needs of the diverse users from the web. It implements abstraction by defining different indicators for directing the user search interests. Results from experiments with Abstraction Focused System approve the success particularly in cases where different users have a defined boundary of the search over WWW.

  2. Performance of the SKB copper/steel canister

    International Nuclear Information System (INIS)

    The performance of the SKB copper/steel canister has been analyzed. The present knowledge of long-term function of the canister is summarized. Radionuclide release calculations for a reference failure scenario and the effect of some variations on release rates are shown. The Features, Events and Processes (FEPs) that are affecting the studied scenarios have been classified according to the 'Rock Engineering Systems' methodology as defined by SKB for the copper/steel canister. Radionuclide release rate is calculated for a reference failure scenario where a small hole in the weld of the outer copper overpack is assumed to exist at the time of deposition. The hole in the copper overpack is assumed to be of a constant size until the inner steel canister looses its mechanical integrity. The steel is assumed to maintain mechanical stability during 5000 years and after this time period the hole through the copper is assumed to be 0.1 m2, which translate to insignificant transport resistance from the canister wall. The release rates for C-14, Sr-90, I-129, Cs-137, Pu-239 and Am-241 are calculated for the reference failure scenario and for a number of variations. The variations include glaciation, only few of the Zircaloy tubes damaged, different canister filling materials, variations in sorption properties of the bentonite clay and different life-time of the inner steel canister. The performance of the canister and near-field, concerning the release rates of the studied radionuclides, is as expected, comparable to the release rates obtained in SKB 91. 11 refs, figs, tabs

  3. ASTRI SST-2M Data Handling and Archiving System

    CERN Document Server

    Antonelli, L Angelo; Lucarelli, Fabrizio; Testa, Vincenzo; Trifoglio, Massimo; Bastieri, Denis; Bulgarelli, Andrea; Capalbi, Milvia; Carosi, Alessandro; Conforti, Vito; Di Paola, Andrea; Gallozzi, Stefano; Gianotti, Fulvio; Perri, Matteo; Tosti, Gino; Rubini, Alda; Vercellone, Stefano

    2013-01-01

    The ASTRI project is the INAF (Italian National Institute for Astrophysics) flagship project developed in the context of the Cherenkov Telescope Array (CTA) international project. ASTRI is dedicated to the realization of the prototype of a Cherenkov small-size dual-mirror telescope (SST-2M) and then to the realization of a mini-array composed of a few of these units. The prototype and all the necessary hardware devices are foreseen to be installed at the Serra La Nave Observing Station (Catania, Italy) in 2014. The upcoming data flow will be properly reduced by dedicated (online and offline) analysis pipelines aimed at providing robust and reliable scientific results (signal detection, sky maps, spectra and light curves) from the ASTRI silicon photo-multipliers camera raw data. Furthermore, a flexible archiving system has being conceived for the storage of all the acquired ASTRI (scientific, calibration, housekeeping) data at different steps of the data reduction up to the final scientific products. In this c...

  4. Query-handling in MLM-based decision support systems.

    Science.gov (United States)

    Arkad, K; Gao, X M; Ahlfeldt, H

    1995-01-01

    Arden Syntax for Medical Logic Modules is a standard specification for creation and sharing of knowledge bases. The standard specification focuses on knowledge that can be represented as a set of independent Medical Logic Modules (MLMs) such as rules, formulas and protocols. The basic functions of an MLM are to retrieve patient data, manipulate the data, come to some decision, and possibly perform an action. All connections to the world outside an MLM are collected in the data-slot of the MLM. The institution specific parts of these connections are inside the notation of curly brackets ([]) to facilitate sharing of MLM between institutions. This paper focuses on some of the problems that occur in relation to Arden Syntax and connections to a patient database such as database queries. Problems related to possibilities of moving one or several module(s) are also discussed, with emphasis on database connections. As an example, an MLM based Decision Support System (DSS) developed at Linköping University is described. PMID:8882561

  5. 76 FR 63714 - Big Spring Rail System, Inc.;Operation Exemption;Transport Handling Specialists, Inc.

    Science.gov (United States)

    2011-10-13

    ... Surface Transportation Board Big Spring Rail System, Inc.;Operation Exemption;Transport Handling Specialists, Inc. Big Spring Rail System, Inc. (BSRS), a noncarrier, has filed a verified notice of exemption....07 in Howard County, Tex., owned by the City of Big Spring, Tex. (City). BSRS will be operating...

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

    International Nuclear Information System (INIS)

    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

  7. ATHLETE: A Cargo-Handling Vehicle for Solar System Exploration

    Science.gov (United States)

    Wilcox, Brian H.

    2011-01-01

    As part of the NASA Exploration Technology Development Program, the Jet Propulsion Laboratory is developing a vehicle called ATHLETE: the All-Terrain Hex-Limbed Extra-Terrestrial Explorer. Each vehicle is based on six wheels at the ends of six multi-degree-of-freedom limbs. Because each limb has enough degrees of freedom for use as a general-purpose leg, the wheels can be locked and used as feet to walk out of excessively soft or other extreme terrain. Since the vehicle has this alternative mode of traversing through or at least out of extreme terrain, the wheels and wheel actuators can be sized for nominal terrain. There are substantial mass savings in the wheel and wheel actuators associated with designing for nominal instead of extreme terrain. These mass savings are comparable-to or larger-than the extra mass associated with the articulated limbs. As a result, the entire mobility system, including wheels and limbs, can be about 25% lighter than a conventional mobility chassis. A side benefit of this approach is that each limb has sufficient degrees-of-freedom to use as a general-purpose manipulator (hence the name "limb" instead of "leg"). Our prototype ATHLETE vehicles have quick-disconnect tool adapters on the limbs that allow tools to be drawn out of a "tool belt" and maneuvered by the limb. A power-take-off from the wheel actuates the tools, so that they can take advantage of the 1+ horsepower motor in each wheel to enable drilling, gripping or other power-tool functions. Architectural studies have indicated that one useful role for ATHLETE in planetary (moon or Mars) exploration is to "walk" cargo off the payload deck of a lander and transport it across the surface. Recent architectural approaches are focused on the concept that the lander descent stage will use liquid hydrogen as a propellant. This is the highest performance chemical fuel, but it requires very large tanks. A natural geometry for the lander is to have a single throttleable rocket engine on

  8. Study and evaluation of innovative fuel handling systems for Sodium cooled Fast Reactors. Single Component Optimization

    International Nuclear Information System (INIS)

    The prototype ASTRID (Advances Sodium Technological Reactor for Industrial Demonstration) sets out to demonstrate the progress made in SFR technology at industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. Among all ASTRID requirements, two are specifically impacting the Fuel Handling Systems (FHS) : the reactor load factor (up to 90%) and the investment costs of the prototype (the ratio of the Fuel Handling System to the total reactor investment cost is estimated to be from 15% to 20%. A large set of fuel handling routes has been investigated. The options considered include in-vessel fuel handling systems (under rotating plugs) and options for the transfer of fuel between the reactor vessel and the external storage. The work performed realized a characterization of solutions, a performance review and an analysis of the advantages and drawbacks of the options compared to a so-called starting reference solution (SRS option) based upon well-known French SFR options or some option already envisaged in French project i.e. EFR reactor. The conclusion of this technological feasibility study is presented later for each option, and a macro-criteria grid analysis has been performed to highlight the innovative options enabled to be pursued for ASTRID. The following options were investigated: a cover core structure (CCS) designed in two independent parts, design of the “Dual Location Rotor” and design of the “Simultaneous Handling of two assemblies”. (author)

  9. System expansion for handling co-products in LCA of sugar cane bio-energy systems

    DEFF Research Database (Denmark)

    Nguyen, T Lan T; Hermansen, John Erik

    2012-01-01

    This study aims to establish a procedure for handling co-products in life cycle assessment (LCA) of a typical sugar cane system. The procedure is essential for environmental assessment of ethanol from molasses, a co-product of sugar which has long been used mainly for feed. We compare system...... abatement scenario, which assumes implementation of substituting bioenergy for fossil-based energy to reduce GHG emissions, combined with a negligible level of emissions from the use stage, keeps the estimate of ethanol life cycle GHG emissions below that of gasoline. Pointing out that indirect land use...... change (ILUC) is a consequence of diverting molasses from feed to fuel, system expansion is the most adequate method when the purpose of the LCA is to support decision makers in weighing the options and consequences. As shown in the sensitivity analysis, an addition of carbon emissions from ILUC worsens...

  10. Efficient Handling of Big Data Volume Using Heterogeneous Distributed File Systems

    OpenAIRE

    Radhakrishnan R; Karthik S

    2014-01-01

    Big Data is the emerging technology of modern world. With the increasing users of online based services the growth of data is tremendous. As the size increases there comes the challenge to handle the large Volume in big data. The objective of this paper is to handle the volume of big data in an efficient way. For this we propose a heterogeneous file system based storage and retrieval method. This uses a three step process, (1) Data nodes with different types of file systems are formed, (2) Th...

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

    International Nuclear Information System (INIS)

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

  12. The on-board data handling system of the AFIS-P mission

    Energy Technology Data Exchange (ETDEWEB)

    Gaisbauer, Dominic; Greenwald, Daniel; Hahn, Alexander; Hauptmann, Philipp; Konorov, Igor; Meng, Lingxin; Paul, Stephan; Poeschl, Thomas [Physics Department E18, Technische Universitaet Muenchen (Germany); Losekamm, Martin [Physics Department E18, Technische Universitaet Muenchen (Germany); Institute of Astronautics, Technische Universitaet Muenchen (Germany); Renker, Dieter [Physics Department E17, Technische Universitaet Muenchen (Germany)

    2014-07-01

    The Antiproton Flux in Space experiment (AFIS) is a novel particle detector comprised of silicon photomultipliers and scintillating plastic fibers. Its purpose is to measure the trapped antiproton flux in low Earth orbit. To test the detector and the data acquisition system, a prototype detector will be flown aboard a high altitude research balloon as part of the REXUS/BEXUS program by the German Aerospace Center (DLR). This talk presents the on-board data handling system and the ground support equipment of AFIS-P. It will also highlight the data handling algorithms developed and used for the mission.

  13. The on-board data handling system of the AFIS-P mission

    International Nuclear Information System (INIS)

    The Antiproton Flux in Space experiment (AFIS) is a novel particle detector comprised of silicon photomultipliers and scintillating plastic fibers. Its purpose is to measure the trapped antiproton flux in low Earth orbit. To test the detector and the data acquisition system, a prototype detector will be flown aboard a high altitude research balloon as part of the REXUS/BEXUS program by the German Aerospace Center (DLR). This talk presents the on-board data handling system and the ground support equipment of AFIS-P. It will also highlight the data handling algorithms developed and used for the mission.

  14. Pilot material handling system for radiation processing of agricultural and medical products

    International Nuclear Information System (INIS)

    A 10 MeV, 10 kW electron LINAC based radiation processing facility is being constructed at Centre for Advanced Technology, Indore for radiation processing of various food products like potatoes, onion, spices, home pack items and medical sterilization. A pilot material handling system has been designed, manufactured, and installed at CAT to verify process parameters viz. conveying speed, dose uniformity, and to study the effect of packing shape and size for radiation processing of different product. This paper describes various features of pilot material handling system. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-15

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

  17. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure

  18. Description of Defense Waste Processing Facility reference waste form and canister. Revision 1

    International Nuclear Information System (INIS)

    The Defense Waste Processing Facility (DWPF) will be located at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1984. The reference waste form is borosilicate glass containing approx. 28 wt % sludge oxides, with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains about 58% SiO2 and 15% B2O3. Leachabilities of SRP waste glasses are expected to approach 10-8 g/m2-day based upon 1000-day tests using glasses containing SRP radioactive waste. Tests were performed under a wide variety of conditions simulating repository environments. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approx. 470 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the sludge and supernate processes. The radionuclide content of the canister is about 177,000 ci, with a radiation level of 5500 rem/h at canister surface contact. The reference canister is fabricated of standard 24-in.-OD, Schedule 20, 304L stainless steel pipe with a dished bottom, domed head, and a combined lifting and welding flange on the head neck. The overall canister length is 9 ft 10 in. with a 3/8-in. wall thickness. The 3-m canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected as an optimum size from glass quality considerations, a logical size for repository handling and to ensure that a filled canister with its double containment shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be compatible with preliminary assessments of repository requirements. 10 references

  19. Fuel handling system of 10 MW high temperature gas cooling reactor based on LabVIEW

    International Nuclear Information System (INIS)

    The field multi-channel signals has been acquired synchronously from 10 MW High temperature gas cooling reactor fuel handling system by DAQ technology. Counting software is developed based on LabVIEW. Its virtual instrument is flexible and user-friendly, and can count fuel-ball exactly. (authors)

  20. SLSF loop handling system. Volume III. AISC code evaluations and analysis of critical attachments

    International Nuclear Information System (INIS)

    SLSF loop handling system was analyzed for deadweight and postulated dynamic loading conditions using a linear elastic static equivalent method of stress analysis. Stress computations of Cradle and critical attachments per AISC Code guidelines are presented. HFEF is credited with in-depth review of initial phase of work

  1. A critical analysis of the X.400 model of message handling systems

    NARCIS (Netherlands)

    Sinderen, van Marten; Dorregeest, Evert

    1988-01-01

    The CCITT X.400 model of store and forward Message Handling Systems (MHS) serves as a common basis for the definition of electronic mail services and protocols both within CCITT and ISO. This paper presents an analysis of this model and its related recommendations from two perspectives. First the co

  2. The system for transportation and handling of radioactive waste packages in the planned Konrad repository

    International Nuclear Information System (INIS)

    The design of the appropriate system for transporting and handling radioactive waste packages in the planned Konrad repository is based, in addition to operational and logistic questions, on results of an incident analysis. The emplacement procedure of the waste packages as well as the essential transfer installations are described and the safety-related aspects are mentioned. ((orig.))

  3. A Review of Active Yaw Control System for Vehicle Handling and Stability Enhancement

    OpenAIRE

    M. K. Aripin; Yahaya Md Sam; Danapalasingam, Kumeresan A.; Kemao Peng; N. Hamzah; Ismail, M. F.

    2014-01-01

    Yaw stability control system plays a significant role in vehicle lateral dynamics in order to improve the vehicle handling and stability performances. However, not many researches have been focused on the transient performances improvement of vehicle yaw rate and sideslip tracking control. This paper reviews the vital elements for control system design of an active yaw stability control system; the vehicle dynamic models, control objectives, active chassis control, and control strategies with...

  4. Interim design status and operational report for semiremote handling fixtures: size reduction system

    International Nuclear Information System (INIS)

    Crushing of HTGR fuel elements is accomplished by a three-stage crushing system consisting of two overhead eccentric jaw crushers, a double-roll crusher, and an oversize reduction system to ensure complete reduction to the desired size. The crushing system is mounted in a special framework which enables gravity flow, eliminates material transport, and minimizes material holdup. The system has been designated UNIFRAME because of the integrated nature of the equipment. This report addresses the demonstration of semiremote maintenance of the crusher in a nonradioactive environment. Although the crusher maintenance system has some remote handling capability inherent in its design, the scope of this initial program is limited to the handling of selected components and allows for manual assistance in certain circumstances. This mode of operation is designated semiremote maintenance and is intended as an effort to gather experience

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

    International Nuclear Information System (INIS)

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

  6. Initial Investigation of Reaction Control System Design on Spacecraft Handling Qualities for Earth Orbit Docking

    Science.gov (United States)

    Bailey, Randall E.; Jackson, E. Bruce; Goodrich, Kenneth H.; Ragsdale, W. Al; Neuhaus, Jason; Barnes, Jim

    2008-01-01

    A program of research, development, test, and evaluation is planned for the development of Spacecraft Handling Qualities guidelines. In this first experiment, the effects of Reaction Control System design characteristics and rotational control laws were evaluated during simulated proximity operations and docking. Also, the influence of piloting demands resulting from varying closure rates was assessed. The pilot-in-the-loop simulation results showed that significantly different spacecraft handling qualities result from the design of the Reaction Control System. In particular, cross-coupling between translational and rotational motions significantly affected handling qualities as reflected by Cooper-Harper pilot ratings and pilot workload, as reflected by Task-Load Index ratings. This influence is masked but only slightly by the rotational control system mode. While rotational control augmentation using Rate Command Attitude Hold can reduce the workload (principally, physical workload) created by cross-coupling, the handling qualities are not significantly improved. The attitude and rate deadbands of the RCAH introduced significant mental workload and control compensation to evaluate when deadband firings would occur, assess their impact on docking performance, and apply control inputs to mitigate that impact.

  7. NDE to Manage Atmospheric SCC in Canisters for Dry Storage of Spent Fuel: An Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cuta, Judith M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Adkins, Harold E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qiao, Hong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Larche, Michael R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, Aaron A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doctor, Steven R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-01

    This report documents efforts to assess representative horizontal (Transuclear NUHOMS®) and vertical (Holtec HI-STORM) storage systems for the implementation of non-destructive examination (NDE) methods or techniques to manage atmospheric stress corrosion cracking (SCC) in canisters for dry storage of used nuclear fuel. The assessment is conducted by assessing accessibility and deployment, environmental compatibility, and applicability of NDE methods. A recommendation of this assessment is to focus on bulk ultrasonic and eddy current techniques for direct canister monitoring of atmospheric SCC. This assessment also highlights canister regions that may be most vulnerable to atmospheric SCC to guide the use of bulk ultrasonic and eddy current examinations. An assessment of accessibility also identifies canister regions that are easiest and more difficult to access through the ventilation paths of the concrete shielding modules. A conceivable sampling strategy for canister inspections is to sample only the easiest to access portions of vulnerable regions. There are aspects to performing an NDE inspection of dry canister storage system (DCSS) canisters for atmospheric SCC that have not been addressed in previous performance studies. These aspects provide the basis for recommendations of future efforts to determine the capability and performance of eddy current and bulk ultrasonic examinations for atmospheric SCC in DCSS canisters. Finally, other important areas of investigation are identified including the development of instrumented surveillance specimens to identify when conditions are conducive for atmospheric SCC, characterization of atmospheric SCC morphology, and an assessment of air flow patterns over canister surfaces and their influence on chloride deposition.

  8. Challenges and Innovative Technologies on Fuel Handling Systems for Future Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Fast Reactors have a unique capability as a sustainable energy source in terms of both utilisation of fissile material for energy production and minimisation of the nuclear waste, due to the hard neutron spectrum. As a result of a screening review of candidate technologies and in the frame of the international forum Generation IV, Sodium Fast Reactors (SFR) are amongst the selected systems to address the sustainability issues with a coherent set of innovative requirements. The guidelines for the definition of such innovative requirements are the Generation IV goals with significant improvements on economy, safety, environment, waste management and proliferation resistance as promising milestone towards a sustainable nuclear energy. CEA, AREVA and EDF have an extensive experience and significant expertise in Sodium cooled Fast Reactors over the past 40 years of R and D and feedback experiments. Some improvements are needed on the SFR to meet the GEN IV goals, and in particular the reduction of investment and operating costs: the Fuel Handling System (FHS) can be considered as an essential step in the reactor design. The reactor refuelling system provides the means of transporting, storing and handling reactor core subassemblies. The system consists of the facilities and equipments 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

  9. Efficient Handling of Big Data Volume Using Heterogeneous Distributed File Systems

    Directory of Open Access Journals (Sweden)

    Radhakrishnan R

    2014-09-01

    Full Text Available Big Data is the emerging technology of modern world. With the increasing users of online based services the growth of data is tremendous. As the size increases there comes the challenge to handle the large Volume in big data. The objective of this paper is to handle the volume of big data in an efficient way. For this we propose a heterogeneous file system based storage and retrieval method. This uses a three step process, (1 Data nodes with different types of file systems are formed, (2 The incoming file size and expected frequency of access is determined, (3 Based on the file size and expected frequency of access a suitable file system is selected. Our method shows that it is much efficient and better than using of single file system for all file sizes and frequency of use, which is shown in our results.

  10. TITLE III EVALUATION REPORT FOR THE MATERIAL AND PERSONNEL HANDLING SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    T. A. Misiak

    1998-05-21

    This Title III Evaluation Report (TER) provides the results of an evaluation that was conducted on the Material and Personnel Handling System. This TER has been written in accordance with the ''Technical Document Preparation Plan for the Mined Geologic Disposal System Title III Evaluation Reports'' (BA0000000-01717-4600-00005 REV 03). The objective of this evaluation is to provide recommendations to ensure consistency between the technical baseline requirements, baseline design, and the as-constructed Material and Personnel Handling System. Recommendations for resolving discrepancies between the as-constructed system, the technical baseline requirements, and the baseline design are included in this report. Cost and Schedule estimates are provided for all recommended modifications.

  11. Initial performance evaluation of major components in the head-end reprocessing solids handling system

    International Nuclear Information System (INIS)

    The General Atomic cold head-end reprocessing pilot plant has been built to demonstrate the feasibility of the proposed commercial reprocessing flowsheet, in particular its integrated operation. This integration is accomplished in part by the solids handling system, which is designed to provide transfer of material at required rates between different steps in the process and to provide the required surge capacity. The major components of the solids handling system have been tested in order to verify or upgrade the design. The components described here are: inlet filters, conveying lines, bunkers, in-bunker filters, blowers, level sensors, feeders, and weigh cells. By and large, the equipment has performed as expected. Feeding of the various materials in the system has received considerable attention, and several improvements were necessary. The system is now equipped to perform its function of serving the needs of the other unit operations in the pilot plant

  12. TITLE III EVALUATION REPORT FOR THE MATERIAL AND PERSONNEL HANDLING SYSTEM

    International Nuclear Information System (INIS)

    This Title III Evaluation Report (TER) provides the results of an evaluation that was conducted on the Material and Personnel Handling System. This TER has been written in accordance with the ''Technical Document Preparation Plan for the Mined Geologic Disposal System Title III Evaluation Reports'' (BA0000000-01717-4600-00005 REV 03). The objective of this evaluation is to provide recommendations to ensure consistency between the technical baseline requirements, baseline design, and the as-constructed Material and Personnel Handling System. Recommendations for resolving discrepancies between the as-constructed system, the technical baseline requirements, and the baseline design are included in this report. Cost and Schedule estimates are provided for all recommended modifications

  13. A comparison of the consequences of different waste handling systems in two Danish communities

    DEFF Research Database (Denmark)

    Grunert, Suzanne C.; Thøgersen, John

    1995-01-01

    Results from a study conducted in two Danish communities with different waste handling systems are reported: Whereas one community introduced in the beginning of 1993 a system of combining economic incentives with structural improvements to promot separation, the other started in spring 1994 a...... cities, and the use of economic incentives were tested. Whereas beliefs influenced attitudes in the expected direction, the consequences of economi incentives for differences in attitudes were less clear....

  14. A critical analysis of the X.400 model of message handling systems

    OpenAIRE

    Sinderen, van, Marten; Dorregeest, Evert

    1988-01-01

    The CCITT X.400 model of store and forward Message Handling Systems (MHS) serves as a common basis for the definition of electronic mail services and protocols both within CCITT and ISO. This paper presents an analysis of this model and its related recommendations from two perspectives. First the concepts of service, protocol and interface are discussed together with their application to this model; second the positioning within ISO's reference model for Open Systems Interconnection (OSI) is ...

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

    International Nuclear Information System (INIS)

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

  16. A Sentinel Approach to Fault Handling in Multi-Agent Systems

    OpenAIRE

    Hägg, Staffan

    1996-01-01

    Fault handling in Multi-Agent Systems (MAS) is not much addressed in current research. Normally, it is considered difficult to address in detail and often well covered by traditional methods, relying on the underlying communication and operating system. In this paper it is shown that this is not necessarily true, at least not with the assumptions on applications we have made. These assumptions are a massive distribution of computing components, a heterogeneous underlying infrastructure (in te...

  17. Characterization of Performance, Robustness, and Behavior Relationships in a Directly Connected Material Handling System

    OpenAIRE

    Anderson, Roger J.

    2006-01-01

    In the design of material handling systems with complex and unpredictable dynamics, conventional search and optimization approaches that are based only on performance measures offer little guarantee of robustness. Using evidence from research into complex systems, the use of behavior-based optimization is proposed, which takes advantage of observed relationships between complexity and optimality with respect to both performance and robustness. Based on theoretical complexity measures, parti...

  18. CAMAC - A modular instrumentation system for data handling. Revised description and specification

    International Nuclear Information System (INIS)

    CAMAC is a modern data handling system in widespread use with on-line digital computers. It is based on a digital highway for data and control. The CAMAC specifications ensures compatibility between equipment from different sources. The revised specification introduces several new features, but is consistent with the previous version (EUR 4100e, 1969). The CAMAC system was specified by European laboratories, through the Esone Committee, and has been endorsed by the USAEC NIM Committee, who have an identical specification (TID-25875)

  19. Canister design concepts for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    As part of its long-term plans for development of a repository for spent fuel (SF) and high level waste (HLW), Nagra is exploring various options for the selection of materials and design concepts for disposal canisters. The selection of suitable canister options is driven by a series of requirements, one of the most important of which is providing a minimum 1000 year lifetime without breach of containment. One candidate material is carbon steel, because of its relatively low corrosion rate under repository conditions and because of the advanced state of overall technical maturity related to construction and fabrication. Other materials and design options are being pursued in parallel studies. The objective of the present study was to develop conceptual designs for carbon steel SF and HLW canisters along with supporting justification. The design process and outcomes result in design concepts that deal with all key aspects of canister fabrication, welding and inspection, short-term performance (handling and emplacement) and long-term performance (corrosion and structural behaviour after disposal). A further objective of the study is to use the design process to identify the future work that is required to develop detailed designs. The development of canister designs began with the elaboration of a number of design requirements that are derived from the need to satisfy the long-term safety requirements and the operational safety requirements (robustness needed for safe handling during emplacement and potential retrieval). It has been assumed based on radiation shielding calculations that the radiation dose rate at the canister surfaces will be at a level that prohibits manual handling, and therefore a hot cell and remote handling will be needed for filling the canisters and for final welding operations. The most important canister requirements were structured hierarchically and set in the context of an overall design methodology. Conceptual designs for SF canisters

  20. User‘s Friendly Interface to the CDF Data Handling System

    Institute of Scientific and Technical Information of China (English)

    F.Ratnikov

    2001-01-01

    The CDF collaboration at the Fermilab Tevatron analyses proton-antiproton interactions at a center-of=mass energy of 2 TeV.during the the collider run starting this year the experiment expects to record 1 Petabyte of data and associated data samples,The Data Handling(DH) system has online and offline components.The DH offline component provides access to the stored data,to stored reconstruction output,to stored Monte-Carlo data samples,and user owned data samples.It serves more than 450 physicists of the collaboration.The extra requirements to the offline component of the Data Handling system are simplicity and convenience for users.More than 50 million events of the CDF Run II data have been already processed using this system.

  1. Design of Central Management & Control Unit for Onboard High-Speed Data Handling System

    Institute of Scientific and Technical Information of China (English)

    LI Yan-qin; JIN Sheng-zhen; NING Shu-nian

    2007-01-01

    The Main Optical Telescope (MOT) is an important payload of the Space Solar Telescope (SST) with various instruments and observation modes. Its real-time data handling and management and control tasks are arduous. Based on the advanced techniques of foreign countries, an improved structure of onboard data handling systems feasible for SST, is proposed. This article concentrated on the development of a Central Management & Control Unit (MCU) based on FPGA and DSP. Through reconfigurating the FPGA and DSP programs, the prototype could perform different tasks.Thus the inheritability of the whole system is improved. The completed dual-channel prototype proves that the system meets all requirements of the MOT. Its high reliability and safety features also meet the requirements under harsh conditions such as mine detection.

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

    International Nuclear Information System (INIS)

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

  3. Design of fuel handling and storage systems in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    The purpose of this Safety Guide is to provide recommendations on the design of fuel handling and storage systems for nuclear power plants. It presents recommendations on how to fulfil the requirements established in the Safety Requirements publication Safety of Nuclear Power Plants: Design. The scope of this Safety Guide is primarily the design of handling and storage systems for fuel assemblies associated with thermal nuclear reactors that are land based. It addresses all stages of fuel handling and storage, which include: the safe receipt of fuel at the nuclear power plant. The storage and inspection of fuel before use. The transfer of fresh fuel into the reactor. The removal of irradiated fuel from the reactor. The reinsertion of irradiated fuel when required. The storage, inspection and repair of the irradiated fuel and its preparation for removal from the reactor pool. The handling of the transport casks. Limited consideration is given to the handling and storage of certain core components, such as reactivity control devices. The recommendations of this Safety Guide also apply to other reactor types as appropriate, such as gas cooled reactors and reactors that are designed for on-load refuelling. Reference provides recommendations on the design of storage facilities for spent fuel, which are not an integral part of an operating nuclear power plant, although such facilities may be located on the same site. Such spent fuel storage facilities provide for the safe storage of spent nuclear fuel after it has been removed from the reactor pool and before it is reprocessed or disposed of as radioactive waste

  4. Estimation of tritium and helium inventory in the tritium handling system in Korea

    International Nuclear Information System (INIS)

    In Korea, the Wolsong Tritium Removal Facility (WTRF) is under construction to reduce the amount of tritium present in the moderator and coolant of the CANDU type Wolsong nuclear power plants. Recently, a study on the tritium handling system for recovery of the tritium collected from the WTRF was started. Some tritium would enter the steel of the container walls and subsequently decay to helium. This helium can deteriorate the mechanical properties of the material of the tritium handling system. To evaluate the tritium and helium inventory in the stainless steel wall of this system, the time-dependent diffusion equation was developed, solved and the results are presented in this paper. These results were compared to previous work that evaluated the tritium inventory in the stainless steel wall of 50-L tritium containers. Tritium and helium concentration profiles and the corresponding inventories were evaluated with respect to the various parameters such as exposure time, temperature, and partial pressure. After 24 years, the helium inventory in the wall of the tritium handling system exceeds the tritium inventory. (authors)

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

    International Nuclear Information System (INIS)

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

  6. The crime prevention and safety system for the radiation handling facilities using network image server

    International Nuclear Information System (INIS)

    A crime prevention and safety system using the image was built for the radiation handling facilities. The system employs an internet network. A specific surveillance monitor would be unnecessary in those facilities where the internet is prepared. The following conclusions were derived as a result of applying this system. (1) Because images were preserved by the digital style, those at the time of accident could be searched easily. (2) The system combined with sensor light and sign deterrent has high ability for warning. (3) A real-time surveillance in the night time can be easily realized. (author)

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

    International Nuclear Information System (INIS)

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

  8. Penentuan Kebijakan Perawatan dan Optimasi Persediaan Suku Cadang pada Coal Handling System PLTU Paiton

    Directory of Open Access Journals (Sweden)

    Fadeli Muhammad F

    2012-09-01

    Full Text Available Fasilitas yang terdapat pada sebuah pembangkit listrik membutuhkan perawatan agar dapat berfungsi  sesuai dengan kapasitasnya. Salah satu fasilitas yang membutuhkan perawatan pada PLTU Paiton adalah fasilitas coal handling system. Perusahaan perlu menerapkan strategi perawatan yang tepat agar biaya perawatan yang dikeluarkan dapat optimal. Permasalahan yang ada pada PLTU Paiton adalah strategi perawatan yang ada masih belum bisa mengatasi kemungkinan kegagalan yang terjadi dan aktivitas perawatan yang dilakukan tidak didukung oleh ketersediaan suku cadang yang dibutuhkan. Hal tersebut menyebabkan biaya perawatan yang dikeluarkan menjadi tidak optimal. Penelitian ini menggunakan metode reliability centered maintenance (RCM II yang dikombinasikan dengan metode evaluasi dari electrical power research institute (EPRI untuk menentukan strategi perawatan yang tepat terhadap coal handling system. Permasalahan persediaan untuk mendukung implementasi penerapan strategi perawatan akan diselesaikan dengan metode probabilistic economic order quantity (EOQ model. Penggunaan metode RCM II yang dikombinasikan dengan metode evaluasi dari EPRI dan penggunaan metode probabilistic EOQ model bertujuan untuk mengoptimalkan biaya perawatan yang dikeluarkan.

  9. An integrated approach for modeling and solving the scheduling problem of container handling systems

    Institute of Scientific and Technical Information of China (English)

    CHEN Lu; XI Li-feng; CAI Jian-guo; BOSTEL Nathalie; DEJAX Pierre

    2006-01-01

    An integrated model is presented to schedule the container handling system. The objective is to improve the cooperation between different types of equipments, and to increase the productivity of the terminal. The problem is formulated as a Hybrid Flow Shop Scheduling problem with precedence constraint, setup times and blocking (HFSS-B). A tabu search algorithm is proposed to solve this problem. The quality and efficiency of the proposed algorithm is analyzed from the computational point of view.

  10. A hybrid framework for the specification of automated material handling systems

    OpenAIRE

    Lau, HYK; Zhao, Y.

    2004-01-01

    This paper presents a hybrid framework that specifies and characterizes the capabilities of generic components in an automated material handling system (AMHS). The framework also provides rules and mechanism for binding these capabilities together so as to facilitate the process of task planning for AMHSs. As a hybrid framework, the formal mathematics of Communicating Sequential Process (CSP) is tightly integrated to the Unified Modeling Language (UML) to provide three important entities, nam...

  11. Material Handling System Design: A Case-Study in Bosch Rexroth Japan

    OpenAIRE

    Akincilar, Sera; Rad, Cameron

    2013-01-01

    In today’s fierce competitive global markets, customers are demanding adjustable lot sizes, shorter lead times, higher quality and flexibility; in short, they want it all. In order to stay competitive in the market, companies need to attain both customer satisfaction and cost reduction in production operations. Material Handling Systems (MHS) is the place to accomplish this goal, since they have a direct impact on production. Therefore, the aim of this study was to design an in-house MHS that...

  12. Canister compatibility with Carlsbad salt

    International Nuclear Information System (INIS)

    No significant reaction was found when candidate canister alloys were heated with salt from Carlsbad, New Mexico, for up to 5000 hours in sealed capsules and for up to 10,000 hours in unsealed capsules at temperatures (80 to 2250C) that bracket the maximum temperature calculated for reference Savannah River Plant (SRP) waste containers at 20-foot spacings in salt. Additional tests were made at 6000C in sealed capsules to characterize reactions that may occur between candidate canister alloys and any component of the salt that is released when decrepitation occurs. Under these extreme conditions there was no significant attack of Type 304L stainless steel. But, there was up to 20-mils attack of the low-carbon steel

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

  15. WALS: A sensor-based robotic system for handling nuclear materials

    International Nuclear Information System (INIS)

    An automated system is being developed for handling large payloads of radioactive nuclear materials in an analytical laboratory. The system uses machine vision and force/torque sensing to provide sensor-based control of the automation system to enhance system safety, flexibility, and robustness and achieve easy remote operation. The automation system also controls the operation of the laboratory measurement systems and the coordination of them with the robotic system. Particular attention has been given to system design features and analytical methods that provide an enhanced level of operational safety. Independent mechanical gripper interlock and too release mechanisms were designed to prevent payload mishandling. An extensive failure modes and effects analysis (FMEA) of the automation system was developed as a safety design analysis tool

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Handling and safety enhancement of race cars using active aerodynamic systems

    Science.gov (United States)

    Diba, Fereydoon; Barari, Ahmad; Esmailzadeh, Ebrahim

    2014-09-01

    A methodology is presented in this work that employs the active inverted wings to enhance the road holding by increasing the downward force on the tyres. In the proposed active system, the angles of attack of the vehicle's wings are adjusted by using a real-time controller to increase the road holding and hence improve the vehicle handling. The handling of the race car and safety of the driver are two important concerns in the design of race cars. The handling of a vehicle depends on the dynamic capabilities of the vehicle and also the pneumatic tyres' limitations. The vehicle side-slip angle, as a measure of the vehicle dynamic safety, should be narrowed into an acceptable range. This paper demonstrates that active inverted wings can provide noteworthy dynamic capabilities and enhance the safety features of race cars. Detailed analytical study and formulations of the race car nonlinear model with the airfoils are presented. Computer simulations are carried out to evaluate the performance of the proposed active aerodynamic system.

  19. Integrated scheduling of a container handling system with simultaneous loading and discharging operations

    Science.gov (United States)

    Li, Chen; Lu, Zhiqiang; Han, Xiaole; Zhang, Yuejun; Wang, Li

    2016-03-01

    The integrated scheduling of container handling systems aims to optimize the coordination and overall utilization of all handling equipment, so as to minimize the makespan of a given set of container tasks. A modified disjunctive graph is proposed and a mixed 0-1 programming model is formulated. A heuristic algorithm is presented, in which the original problem is divided into two subproblems. In the first subproblem, contiguous bay crane operations are applied to obtain a good quay crane schedule. In the second subproblem, proper internal truck and yard crane schedules are generated to match the given quay crane schedule. Furthermore, a genetic algorithm based on the heuristic algorithm is developed to search for better solutions. The computational results show that the proposed algorithm can efficiently find high-quality solutions. They also indicate the effectiveness of simultaneous loading and discharging operations compared with separate ones.

  20. An Optimized Small Tissue Handling System for Immunohistochemistry and In Situ Hybridization.

    Science.gov (United States)

    Anthony, Giovanni; Lee, Ju-Ahng

    2016-01-01

    Recent development in 3D printing technology has opened an exciting possibility for manufacturing 3D devices on one's desktop. We used 3D modeling programs to design 3D models of a tissue-handling system and these models were "printed" in a stereolithography (SLA) 3D printer to create precision histology devices that are particularly useful to handle multiple samples with small dimensions in parallel. Our system has been successfully tested for in situ hybridization of zebrafish embryos. Some of the notable features include: (1) A conveniently transferrable chamber with 6 mesh-bottomed wells, each of which can hold dozens of zebrafish embryos. This design allows up to 6 different samples to be treated per chamber. (2) Each chamber sits in a well of a standard 6-well tissue culture plate. Thus, up to 36 different samples can be processed in tandem using a single 6 well plate. (3) Precisely fitting lids prevent solution evaporation and condensation, even at high temperatures for an extended period of time: i.e., overnight riboprobe hybridization. (4) Flat bottom mesh maximizes the consistent treatment of individual tissue samples. (5) A magnet-based lifter was created to handle up to 6 chambers (= 36 samples) in unison. (6) The largely transparent resin aids in convenient visual inspection both with eyes and using a stereomicroscope. (7) Surface engraved labeling enables an accurate tracking of different samples. (8) The dimension of wells and chambers minimizes the required amount of precious reagents. (9) Flexible parametric modeling enables an easy redesign of the 3D models to handle larger or more numerous samples. Precise dimensions of 3D models and demonstration of how we use our devices in whole mount in situ hybridization are presented. We also provide detailed information on the modeling software, 3D printing tips, as well as 3D files that can be used with any 3D printer. PMID:27489962

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

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

  3. 单柄DP控制系统设计%Design of Single Handle DP Control System

    Institute of Scientific and Technical Information of China (English)

    陈相友

    2012-01-01

    This paper presents the solution of a single handle DP bridge control system used in full rotation tugboat with the control scheme of the hardware and software system.%本文提出了一种对全回转拖轮单柄DP驾驶台控制系统设计方案。本方案对全回转拖轮单柄DP驾驶台控制系统的硬件系统、软件系统进行了系统设计。

  4. Protection of Occupational Users of Fuel Handling System against Nuclear Radiation

    International Nuclear Information System (INIS)

    The reactor hall of ETRR-1 contains the old spent fuel pool whose capacity is only sixty units. Therefore, a new spent fuel storage pool has been constructed which can contains all the units. To protect the occupational operators during the transportation of the spent fuel to the new pool, a safe shielded fuel handling system is required. In this system, a spent fuel basket is lifted from the pool by the use of a shielded metallic cask. The dose rate has been calculated at the outer cask surface for different periods of cooling time and compared with the value of the dose operational limit (10μSv/hr)

  5. Multi-Canister Overpack (MCO) Topical Report

    International Nuclear Information System (INIS)

    In February 1995, the US Department of Energy (DOE) approved the Spent Nuclear Fuel (SNF) Project's ''Path Forward'' recommendation for resolution of the safety and environmental concerns associated with the deteriorating SNF stored in the Hanford Site's K Basins (Hansen 1995). The recommendation included an aggressive series of projects to design, construct, and operate systems and facilitates to permit the safe retrieval, packaging, transport, conditions, and interim storage of the K Basins' SNF. The facilities are the Cold VAcuum Drying Facility (CVDF) in the 100 K Area of the Hanford Site and the Canister Storage building (CSB) in the 200 East Area. The K Basins' SNF is to be cleaned, repackaged in multi-canister overpacks (MCOs), removed from the K Basins, and transported to the CVDF for initial drying. The MCOs would then be moved to the CSB and weld sealed (Loscoe 1996) for interim storage (about 40 years). One of the major tasks associated with the initial Path Forward activities is the development and maintenance of the safety documentation. In addition to meeting the construction needs for new structures, the safety documentation for each must be generated

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. TRUPACT-I, a contact-handled transuranic waste transportation system

    International Nuclear Information System (INIS)

    In the late 1970's, the Department of Energy (DOE) initiated a program to develop an efficient, safe, reliable and cost effective transportation system for the carriage of contact-handled transuranic waste within the DOE complex. The system developed (TRUPACT-I) consists of two major assemblies: the containment system and the outer protective structure. The containment system prevents the release of TRU nuclides in excess of regulatory limits. The outer protective structure protects against the structural and thermal loadings imposed during normal and accident conditions. A fully-loaded truck version of TRUPACT-I will weigh 22.7 tonnes (50,000 lbs) and has a payload of 7.7 tonnes (17,000 lb). TRUPACT-I has been designed to accommodate a variety of material handling equipment such as lift trucks, powered roller conveyors and possibly air pallet systems. The prototype will be fabricated and tested during late FY 84. A three unit minifleet will be operated during FY 85 and FY 86 to gain operational experience

  9. 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.

  10. A transputer based on-board data handling system for small satellites

    Science.gov (United States)

    Ciecior, F.; Arens-Fischer, W.; Iglseder, H.; Backhus, E.; Rath, H. J.

    The on-board data handling system described in this paper is a concurrent real-time data processing system for small satellites. The system is modular; its functional units are strictly separated and implemented using autonomous, independent operating components. The system control software is modular, and its functional units are strictly separated and divided into a configuration layer, a communication layer, and an application layer, each of which is dynamically loaded by the underlying layer. The configuration layer translates the physical network topology into a logical one. The communication layer uses this information for automatic and transparent communication between processes running in parallel on different hardware modules. The application layer comprises an application process interface (API) and all application processes. The API performs file management, resource control, command processing, and communication control functions. This interface in combination with the two underlying software layers implements a simple satellite operating system.

  11. Handle系统的发展及应用*%The Development and Application of Handle System

    Institute of Scientific and Technical Information of China (English)

    郭晓峰; 孙洵

    2013-01-01

    Handle系统作为一种通用的名称服务系统,能够为网络中的数字对象提供的永久标识、动态链接和安全管理等基础服务。文章重点介绍Handle系统的发展及应用情况,从数字图书馆、数字内容管理、数字出版、数字博物馆、远程教育到科学数据管理与网格计算、数字权益管理、信息安全管理与隐私保护等领域。近年来随着互联网的发展,以及物联网等新技术的兴起,Handle系统获得了更加广阔的应用前景。文章最后介绍了Handle系统作为联合国ITU推荐的互联网下一代链接技术,在技术上及管理方式上的最新发展情况,并对其应用前景进行展望。%As a generic name service system, Handle System has provided global services for digital objects over Internet, including persistent identifier, dynamic linking, security management, and other basic services. This article focuses on the developments and applications of the Handle System, from the areas as Digital Library, Digital Contents Management, Digital Publishing, Digital Museum, E-learning to Science Data Management and Grid Computing, Digital Rights Management, Information Security Management and Privacy Protection, etc. Recently, with the rise of new technologies of the Internet and the Internet of Things, more broad application prospects of Handle system are emerging. At the last section, the article introduces the state of the art in the technical and management of Handle System since it has been chosen by ITU as the next generation of the Internet linking technology, and final y prospects its future applications.

  12. Comparison of control systems applied to the handling of radioactive reactor components

    International Nuclear Information System (INIS)

    The first generation of nuclear power stations have individual reactors each incorporating complete facilities for servicing components and refuelling. In the later designs, each power station has two reactors which are connected by a central block. This central block contains one set of facilities to service both reactors, but to improve the station capability, some of these are to be replicated. The central block incorporates a hoist well which was used during construction for the accessing of complete components. On completion of this work, the physical size of the hoist well is such as to permit the incorporation of additional facilities if these are shown to be operationally and economically desirable. Since a number of years of power operation has elapsed, the advantages of back-fitting to existing fuel-handling facilities has been illustrated. Since the mechanical arrangements and operating procedures are substantially similar for both the original and new handling facilities, the paper will illustrate the control systems provided for each. The configuration of the system is arranged to have two channels of control which complies with the current standard requirements in the United Kingdom. These requirements are more stringent than when the existing facility was designed and constructed, as described in the relevant sections of the paper. The new system has been designed and is being manufactured to comply with the Central Electricity Generating Board standard for nuclear fuel route interlock and control systems. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  14. The irradiated fuel transport and handling system used in E.D.F. nuclear installations

    International Nuclear Information System (INIS)

    Since the problem of removal of irradiated fuels is highly important, the solution adopted by the PWR-Plants ensemble of the EDF has been the object of an in-depth study. First, a certain number of directions were defined for the study to be affected: the adoption of a means of removal which is unique for each type of power level (900 and 1300 MW); decrease of the risks of incident by reducing the number of handling steps and by simplifying the operations; decreasing the irradiation of personnel to a minimum by employing systems of protection, decontamination, and appropriate intervention, and by the smallest possible number of removals; and transport by railroad over long distances with the possibility of exceptional removal or short distance transport by motor vehicle. The restraints involved with the following have been considered: the conception of installation and the means of handling, at the station as well as at the recycling plant; handling security in the fuel building which is either active (redundant bridge), or passive (dash pot, concrete cells) or the two; the conception and rules of approbation of irradiated fuel containers, as well as the available possibilities on the market; and the means of transport (weight and speed limits). The result of the study is, for normal transport, a 100 ton container with a 12 element capacity. The delicate tilting manoeuvre must be performed with a specialized installation. Protection against contamination is to be obtained by an hermetic system which encloses the container. An automatic hydrolaser system is envisaged for principal decontamination. The adoption of these principles will thus permit the EDF to evacuate, and the recycler to receive, the irradiated fuel under the most favorable conditions for the public and for the personnel of the installations involved

  15. Synergy of automation and human action in fuel handling control system of 500 MWe PHWR

    International Nuclear Information System (INIS)

    On-line refueling process is carried out in PHWRs on a routine basis. Each refueling cycle involves a large number of sequential operations to be executed safely and satisfactorily and hence is time consuming. Operations on the system are performed as per the detailed prescribed procedures. This necessitates introduction of automation of these operations; however the extent of automation has to be based on consideration of professional motivation and psychological well-being of the operator. This paper describes the nature of fuel handling operations and how man and machine work together in a complementary manner to achieve the objectives. (author). 2 refs., 7 figs., 1 tab

  16. Handling information and knowledge in the FORUM hypermedia authoring system: application to uterine magnetic resonance imaging.

    Science.gov (United States)

    Soula, G; Moulin, G; Delquie, P; Bartoli, J M; Fieschi, M; Kasbarian, M

    1995-01-01

    Hypertext and hypermedia are used to access loosely structured information. The lack of conceptual model for hypertext application makes it difficult to represent the mental model of the author. Users are often lost in the information and the objective of the author is not achieved. In this paper, we describe an approach to handle information and knowledge in the FORUM hypermedia authoring system. It allows the author to create and maintain the hypertext more easily and facilitates the navigation of the reader. An application for learning uterine magnetic resonance imaging is also presented. PMID:8591409

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

    International Nuclear Information System (INIS)

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

  18. Development of the RAW and SNF handling system in the Federal Atomic Energy Agency and in the Russian Federation

    International Nuclear Information System (INIS)

    The experience the Russian Federation has gained over the period longer than half a century in handling radioactive wastes (RAW), including their burial in geological formations, is quite unique. These guidelines state the necessity of establishing an integrated state RAW and spent nuclear fuel (SNF) handling system, ensuring safe, effective, efficient, and socially-acceptable development of using nuclear power to solve issues of national security in all its aspects and utilising secondary recovered and re-used raw materials. This system is increasing safety of handling radioactive materials at all stages of their life cycle. (author)

  19. Material handling systems for use in glovebox lines: A survey of Department of Energy facility experience

    International Nuclear Information System (INIS)

    The Nuclear Weapons Complex Reconfiguration Study has recommended that a new manufacturing facility be constructed to replace the Rocky Flats Plant. In the new facility, use of an automated material handling system for movement of components would reduce both the cost and radiation exposure associated with production and maintenance operations. Contamination control would be improved between process steps through the use of airlocks and portals. Part damage associated with improper transport would be reduced, and accountability would be increased. In-process workpieces could be stored in a secure vault, awaiting a request for parts at a production station. However, all of these desirable features rely on the proper implementation of an automated material handling system. The Department of Energy Weapons Production Complex has experience with a variety of methods for transporting discrete parts in glovebox lines. The authors visited several sites to evaluate the existing technologies for their suitability for the application of plutonium manufacturing. Technologies reviewed were Linear motors, belt conveyors, roller conveyors, accumulating roller conveyors, pneumatic transport, and cart systems. The sites visited were The Idaho National Engineering laboratory, the Hanford Site, and the Rocky Flats Plant. Linear motors appear to be the most promising technology observed for the movement of discrete parts, and further investigation is recommended

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

    International Nuclear Information System (INIS)

    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

  1. Analysis of water from K west basin canisters (second campaign)

    Energy Technology Data Exchange (ETDEWEB)

    Trimble, D.J., Fluor Daniel Hanford

    1997-03-06

    Gas and liquid samples have been obtained from a selection of the approximately 3,820 spent fuel storage canisters in the K West Basin. The samples were taken to characterize the contents of the gas and water in the canisters. The data will provide source term information for two subprojects of the Spent Nuclear Fuel Project (SNFP) (Fulton 1994): the K Basins Integrated Water Treatment System subproject (Ball 1996) and the K Basins Fuel Retrieval System subproject (Waymire 1996). The barrels of ten canisters were sampled in 1995, and 50 canisters were sampled in a second campaign in 1996. The analysis results for the gas and liquid samples of the first campaign have been reported (Trimble 1995a; Trimble 1995b; Trimble 1996a; Trimble 1996b). An analysis of cesium-137 (137CS ) data from the second campaign samples was reported (Trimble and Welsh 1997), and the gas sample results are documented in Trimble 1997. This report documents the results of all analytes of liquid samples from the second campaign.

  2. Options on fuel handling systems for future sodium cooled fast reactors

    International Nuclear Information System (INIS)

    The definition of innovative requirements for SFR (Sodium cooled Fast Reactors) are guided by significant improvements on economy, safety, environment, waste management and proliferation resistance. CEA, AREVA and EDF have achieved an extensive experience and significant expertise in SFR over the past 40 years of research/development and feedback on experiments. Some improvements are needed on the SFR to reach the GEN IV objectives. The reactor refueling system provides the ways of transporting, storing and handling reactor core fuel assemblies. The Fuel Handling System (FHS) impacts directly the general design of the reactor vessel, the reactor building and the nuclear island, their construction cost and the 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. In addition, the feasibility of the Whole Core Discharge (WCD) is studied and consequences in terms of usability and availability are detailed. The various studies carried out in the framework of the Research-Development programs and for the construction of prototypes have led to the definition of two types of FHS: 1) A sodium route: this route is very interesting for operation flexibility but requires an external storage (filled with sodium) which penalize the investment cost. 2) A gas route: this route allows a reduction of the investment cost but operation gets more complicated and possibilities of fast Whole Core Discharge are limited. In the framework of the tripartite project (CEA, EDF and AREVA), we have then considered a 'Mixed' route mainly answering to the following objectives: - Reduction of the investment cost compared to the sodium route - Ability to make a Whole Core Discharge within a few months. After dealing with main constraints and assumptions linked to the SFR Fuel Handling System, we will detail in this paper the maximum fuel assemblies power acceptable for the

  3. Inspection of disposal canisters components

    International Nuclear Information System (INIS)

    This report presents the inspection techniques of disposal canister components. Manufacturing methods and a description of the defects related to different manufacturing methods are described briefly. The defect types form a basis for the design of non-destructive testing because the defect types, which occur in the inspected components, affect to choice of inspection methods. The canister components are to nodular cast iron insert, steel lid, lid screw, metal gasket, copper tube with integrated or separate bottom, and copper lid. The inspection of copper material is challenging due to the anisotropic properties of the material and local changes in the grain size of the copper material. The cast iron insert has some acoustical material property variation (attenuation, velocity changes, scattering properties), which make the ultrasonic inspection demanding from calibration point of view. Mainly three different methods are used for inspection. Ultrasonic testing technique is used for inspection of volume, eddy current technique, for copper components only, and visual testing technique are used for inspection of the surface and near surface area

  4. ECG Sensor Verification System with Mean-Interval Algorithm for Handling Sport Issue

    Directory of Open Access Journals (Sweden)

    Kuo-Kun Tseng

    2016-01-01

    Full Text Available With the development of biometric verification, we proposed a new algorithm and personal mobile sensor card system for ECG verification. The proposed new mean-interval approach can identify the user quickly with high accuracy and consumes a small amount of flash memory in the microprocessor. The new framework of the mobile card system makes ECG verification become a feasible application to overcome the issues of a centralized database. For a fair and comprehensive evaluation, the experimental results have been tested on public MIT-BIH ECG databases and our circuit system; they confirm that the proposed scheme is able to provide excellent accuracy and low complexity. Moreover, we also proposed a multiple-state solution to handle the heat rate changes of sports problem. It should be the first to address the issue of sports in ECG verification.

  5. Integrating plant-internal and plant-external information systems for optimal handling of nuclear emergencies

    International Nuclear Information System (INIS)

    The handling of nuclear emergencies is a complex task, which may involve rescue personnel at the emergency site as well as personnel residing at significant distance from the disaster area. This paper will focus on integration of support systems for use in emergencies arising from nuclear power plant accidents and will cover both plant-internal and plant-external aspects. The different decision centers being involved in management of nuclear accidents, the plant control room, the technical support center and the regional and national emergency centers, are briefly described and focus is placed on the utilization of a common information base, containing basic information being presented at an inter-center level. The utilization of advanced information systems for decision support as well as national emergency management systems is discussed in an attempt to optimize their use in an integrated information network

  6. Study and Handling Methods of Power IGBT Module Failures in Power Electronic Converter Systems

    DEFF Research Database (Denmark)

    Choi, Uimin; Blaabjerg, Frede; Lee, Kyo-Beum

    2015-01-01

    availability in different applications. This paper presents an overview of the major failure mechanisms of IGBT modules and their handling methods in power converter systems improving reliability. The major failure mechanisms of IGBT modules are presented first, and methods for predicting lifetime and......Power electronics plays an important role in a wide range of applications in order to achieve high efficiency and performance. Increasing efforts are being made to improve the reliability of power electronics systems to ensure compliance with more stringent constraints on cost, safety, and...... estimating the junction temperature of IGBT modules are then discussed. Subsequently, different methods for detecting open- and short-circuit faults are presented. Finally, fault-tolerant strategies for improving the reliability of power electronic systems under field operation are explained and compared in...

  7. A Review of Active Yaw Control System for Vehicle Handling and Stability Enhancement

    Directory of Open Access Journals (Sweden)

    M. K. Aripin

    2014-01-01

    Full Text Available Yaw stability control system plays a significant role in vehicle lateral dynamics in order to improve the vehicle handling and stability performances. However, not many researches have been focused on the transient performances improvement of vehicle yaw rate and sideslip tracking control. This paper reviews the vital elements for control system design of an active yaw stability control system; the vehicle dynamic models, control objectives, active chassis control, and control strategies with the focus on identifying suitable criteria for improved transient performances. Each element is discussed and compared in terms of their underlying theory, strengths, weaknesses, and applicability. Based on this, we conclude that the sliding mode control with nonlinear sliding surface based on composite nonlinear feedback is a potential control strategy for improving the transient performances of yaw rate and sideslip tracking control.

  8. ITER Upper Port Plug handling cask system assessment and design proposals

    International Nuclear Information System (INIS)

    The current design of the ITER cask for Upper Port Plugs has been evaluated. Careful reduction of the number of mechanical degrees of freedom is an opportunity to relax the tolerances in the design, resulting in cost reduction and reliability increase. A new kinematical design for the tractor module has a higher stiffness to weight ratio, reduces actuator forces by a factor four and minimizes cross-talk between lift and rotation motion. Non-cantilevered handling is recommended to reduce wheel loads on the tractor by a factor six and to simplify guidance. At the system level the tubular guide (TG) is proposed, a semi-permanent 3.5 m long tube which is an extension of the Upper Port. Cask docking is simplified and the risk of the cask tilting is prevented. Redesigning the system concept is recommended and the TG looks promising. Since a system level redesign impacts the external interfaces, overall feasibility has to be investigated.

  9. Transportation considerations related to waste forms and canisters for Defense TRU wastes

    International Nuclear Information System (INIS)

    This report identifies and discusses the considerations imposed by transportation on waste forms and canisters for contact-handled, solid transuranic wastes from the US Department of Energy (DOE) activities. The report reviews (1) the existing raw waste forms and potential immobilized waste forms, (2) the existing and potential future DOE waste canisters and shipping containers, (3) regulations and regulatory trends for transporting commercial transuranic wastes on the ISA, (4) truck and rail carrier requirements and preferences for transporting the wastes, and (5) current and proposed Type B external packagings for transporting wastes

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

    International Nuclear Information System (INIS)

    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)

  11. Status of the JET active gas handling system and plans for tritium operation

    International Nuclear Information System (INIS)

    The Joint European Torus (JET) carried out the first preliminary experiment with a deuterium-tritium plasma in 1991. This utilised an on-site inventory of 0.25g. The future experimental programme for the JET machine includes two discrete phases using plasmas fuelled by deuterium and tritium. The first of these, in mid-1996, will generate around 2 x 1020 neutrons and require a site inventory of a few grams of tritium. The second is proposed to take place in 1999 if an extension to the JET project from 1996 is granted. This will require a few tens of grams of tritium and will generate up to 5 x 1021 neutrons. The JET Active Gas Handling System has been constructed to enable tritium to be recovered from the plasma exhaust and stored for re-injection. The design also minimises tritium discharges to the environment. It is currently being commissioned to meet the above programme and has been modified to take into account a new requirement for operation over extended periods during maintenance and D-D operation with tritium contaminated plasma exhaust. Commissioning of the Active Gas Handling System consists of inactive, trace tritium (approx.40TBq) and full tritium (<3g) phases. The experience and main results of inactive commissioning are presented and the status of tritium commissioning is reviewed. 17 refs., 1 fig

  12. Investigating the intrinsic cleanliness of automated handling designed for EUV mask pod-in-pod systems

    Science.gov (United States)

    Brux, O.; van der Walle, P.; van der Donck, J. C. J.; Dress, P.

    2011-11-01

    Extreme Ultraviolet Lithography (EUVL) is the most promising solution for technology nodes 16nm (hp) and below. However, several unique EUV mask challenges must be resolved for a successful launch of the technology into the market. Uncontrolled introduction of particles and/or contamination into the EUV scanner significantly increases the risk for device yield loss and potentially scanner down-time. With the absence of a pellicle to protect the surface of the EUV mask, a zero particle adder regime between final clean and the point-of-exposure is critical for the active areas of the mask. A Dual Pod concept for handling EUV masks had been proposed by the industry as means to minimize the risk of mask contamination during transport and storage. SuSS-HamaTech introduces MaskTrackPro InSync as a fully automated solution for the handling of EUV masks in and out of this Dual Pod System and therefore constitutes an interface between various tools inside the Fab. The intrinsic cleanliness of each individual handling and storage step of the inner shell (EIP) of this Dual Pod and the EUV mask inside the InSync Tool has been investigated to confirm the capability for minimizing the risk of cross-contamination. An Entegris Dual Pod EUV-1000A-A110 has been used for the qualification. The particle detection for the qualification procedure was executed with the TNO's RapidNano Particle Scanner, qualified for particle sizes down to 50nm (PSL equivalent). It has been shown that the target specification of mitigate the risk of material abrasion.

  13. Drying behavior of K-East canister sludge

    International Nuclear Information System (INIS)

    A series of tests were conducted by Pacific Northwest National Laboratory to evaluate the drying behavior of sludge taken from the Hanford K-East Basin storage canisters. Some of the components of K-Basin sludge, such as oxides of uranium and its hydrates, could be associated with the spent nuclear fuel that will ultimately be loaded into Multi-Canister Overpacks (MCOs) and transferred to interim dry storage on the Hanford Site. The materials sealed in the MCOs must be compatible with the storage facility safety basis and the design accident analyses. Understanding the drying behavior of hydrates that may be formed by the reaction of uranium oxides (corrosion products) and water will help ensure these criteria are addressed. Drying measurements of sludge samples collected from K-East Basin canisters showed the water content (physically plus chemically bound) to range between 5 wt% and 75 wt%. Uranium oxide hydrates, the main source of gaseous products that can pressurize the MCOs during storage, constituted about 3 wt% to 15 wt% of the total water content of the initial weight. Most of the physically bound water was assumed to be released from the samples at ambient temperature when the system was pumped down to vacuum conditions of about 40 mTorr. The period for release of most free water in the K-East canister sludge was about 24 hours

  14. A Sample Handling System for Mars Sample Return - Design and Status

    Science.gov (United States)

    Allouis, E.; Renouf, I.; Deridder, M.; Vrancken, D.; Gelmi, R.; Re, E.

    2009-04-01

    A mission to return atmosphere and soil samples form the Mars is highly desired by planetary scientists from around the world and space agencies are starting preparation for the launch of a sample return mission in the 2020 timeframe. Such a mission would return approximately 500 grams of atmosphere, rock and soil samples to Earth by 2025. Development of a wide range of new technology will be critical to the successful implementation of such a challenging mission. Technical developments required to realise the mission include guided atmospheric entry, soft landing, sample handling robotics, biological sealing, Mars atmospheric ascent sample rendezvous & capture and Earth return. The European Space Agency has been performing system definition studies along with numerous technology development studies under the framework of the Aurora programme. Within the scope of these activities Astrium has been responsible for defining an overall sample handling architecture in collaboration with European partners (sample acquisition and sample capture, Galileo Avionica; sample containment and automated bio-sealing, Verhaert). Our work has focused on the definition and development of the robotic systems required to move the sample through the transfer chain. This paper presents the Astrium team's high level design for the surface transfer system and the orbiter transfer system. The surface transfer system is envisaged to use two robotic arms of different sizes to allow flexible operations and to enable sample transfer over relatively large distances (~2 to 3 metres): The first to deploy/retract the Drill Assembly used for sample collection, the second for the transfer of the Sample Container (the vessel containing all the collected samples) from the Drill Assembly to the Mars Ascent Vehicle (MAV). The sample transfer actuator also features a complex end-effector for handling the Sample Container. The orbiter transfer system will transfer the Sample Container from the capture

  15. Multi-Canister overpack internal HEPA filters

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, K.E.

    1998-11-03

    The rationale for locating a filter assembly inside each Multi-Canister Overpack (MCO) rather than include the filter in the Cold Vacuum Drying (CVD) process piping system was to eliminate the potential for contamination to the operators, processing equipment, and the MCO. The internal HEPA filters provide essential protection to facility workers from alpha contamination, both external skin contamination and potential internal depositions. Filters installed in the CVD process piping cannot mitigate potential contamination when breaking the process piping connections. Experience with K-Basin material has shown that even an extremely small release can result in personnel contamination and costly schedule disruptions to perform equipment and facility decontamination. Incorporating the filter function internal to the MCO rather than external is consistent with ALARA requirements of 10 CFR 835. Based on the above, the SNF Project position is to retain the internal HEPA filters in the MCO design.

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

    International Nuclear Information System (INIS)

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

  17. A Data Envelopment Analysis Model for Selecting Material Handling System Designs

    Science.gov (United States)

    Liu, Fuh-Hwa Franklin; Kuo, Wan-Ting

    The material handling system under design is an unmanned job shop with an automated guided vehicle that transport loads within the processing machines. The engineering task is to select the design alternatives that are the combinations of the four design factors: the ratio of production time to transportation time, mean job arrival rate to the system, input/output buffer capacities at each processing machine, and the vehicle control strategies. Each of the design alternatives is simulated to collect the upper and lower bounds of the five performance indices. We develop a Data Envelopment Analysis (DEA) model to assess the 180 designs with imprecise data of the five indices. The three-ways factorial experiment analysis for the assessment results indicates the buffer capacity and the interaction of job arrival rate and buffer capacity affect the performance significantly.

  18. Evaluation of a pilot fish handling system at Bruce NGS 'A'

    International Nuclear Information System (INIS)

    A pilot fish recovery system using a Hidrostal fish pump was tested in the Bruce NGS 'A' forebay during June, 1984. Despite low forebay fish concentrations, the system was capable of capturing 97,000 alewife/day (3900 kg) if operated continuously. Post-pumping survival averaged 97%. It is estimated that a single pump could handle alewife runs in the 40,000 to 70,000 kg range, but multiple pumps or a single larger pump would be required to assure station protection from the largest runs (>100,000 kg). Results indicate that tank/trailer return of pumped fish is feasible, but other alternatives for returning fish to Lake Huron are also being considered

  19. Object-oriented data handling and OODB operation of LHD mass data acquisition system

    Energy Technology Data Exchange (ETDEWEB)

    Nakanishi, H. E-mail: nakanisi@nifs.ac.jp; Emoto, M.; Kojima, M.; Ohsuna, M.; Komada, S

    2000-08-01

    The new data acquisition system of large helical device (LHD) diagnostics, i.e. LABCOM system, has successfully started its operation in March 1998. It has a simple but massive parallel-processing (MPP) structure by means of multiple PC/Windows NT environment, and the most significant methodology adopted for it is the object-oriented (OO) data handling through the whole system. The functions and data substances of the acquisition system are described in autonomous objects with the corresponding C++ class definitions. The object-oriented database management system (ODBMS) will be the only solution to provide a vast and virtual storage space for storing an enormous number of archiving data objects. Commercial ODBMS product 'O2' are installed on each diagnostic acquisition computer. Practical O2 investigations showed 300-400 kB/s as the data storing rate, whereas the data transfer rate from CAMAC digitizers to the computer is up to 700 kB/s in this system. Applying the GNU project's 'zlib' compression library for the data size reduction compensates this rate gap. Through the first and second ({approx} no. 7132) LHD experimental campaigns, the LABCOM system acquired about 400 GB raw data, with maximum 120 MB per shot. These experiences proved that OO technology has great promise for the next generation of the data acquisition and storage system in fusion research experiments.

  20. A Globally Distributed System for Job, Data, and Information Handling for High Energy Physics

    Energy Technology Data Exchange (ETDEWEB)

    Garzoglio, Gabriele

    2005-12-01

    The computing infrastructures of the modern high energy physics experiments need to address an unprecedented set of requirements. The collaborations consist of hundreds of members from dozens of institutions around the world and the computing power necessary to analyze the data produced surpasses already the capabilities of any single computing center. A software infrastructure capable of seamlessly integrating dozens of computing centers around the world, enabling computing for a large and dynamical group of users, is of fundamental importance for the production of scientific results. Such a computing infrastructure is called a computational grid. The SAM-Grid offers a solution to these problems for CDF and DZero, two of the largest high energy physics experiments in the world, running at Fermilab. The SAM-Grid integrates standard grid middleware, such as Condor-G and the Globus Toolkit, with software developed at Fermilab, organizing the system in three major components: data handling, job handling, and information management. This dissertation presents the challenges and the solutions provided in such a computing infrastructure.

  1. The Preemptive Stocker Dispatching Rule of Automatic Material Handling System in 300 mm Semiconductor Manufacturing Factories

    Science.gov (United States)

    Wang, C. N.; Lin, H. S.; Hsu, H. P.; Wang, Yen-Hui; Chang, Y. P.

    2016-04-01

    The integrated circuit (IC) manufacturing industry is one of the biggest output industries in this century. The 300mm wafer fabs is the major fab size of this industry. The automatic material handling system (AMHS) has become one of the most concerned issues among semiconductor manufacturers. The major lot delivery of 300mm fabs is used overhead hoist transport (OHT). The traffic jams are happened frequently due to the wide variety of products and big amount of OHTs moving in the fabs. The purpose of this study is to enhance the delivery performance of automatic material handling and reduce the delay and waiting time of product transportation for both hot lots and normal lots. Therefore, this study proposes an effective OHT dispatching rule: preemptive stocker dispatching (PSD). Simulation experiments are conducted and one of the best differentiated preemptive rule, differentiated preemptive dispatching (DPD), is used for comparison. Compared with DPD, The results indicated that PSD rule can reduce average variable delivery time of normal lots by 13.15%, decreasing average variable delivery time of hot lots by 17.67%. Thus, the PSD rule can effectively reduce the delivery time and enhance productivity in 300 mm wafer fabs.

  2. A Globally Distributed System for Job, Data, and Information Handling for High Energy Physics

    International Nuclear Information System (INIS)

    The computing infrastructures of the modern high energy physics experiments need to address an unprecedented set of requirements. The collaborations consist of hundreds of members from dozens of institutions around the world and the computing power necessary to analyze the data produced surpasses already the capabilities of any single computing center. A software infrastructure capable of seamlessly integrating dozens of computing centers around the world, enabling computing for a large and dynamical group of users, is of fundamental importance for the production of scientific results. Such a computing infrastructure is called a computational grid. The SAM-Grid offers a solution to these problems for CDF and DZero, two of the largest high energy physics experiments in the world, running at Fermilab. The SAM-Grid integrates standard grid middleware, such as Condor-G and the Globus Toolkit, with software developed at Fermilab, organizing the system in three major components: data handling, job handling, and information management. This dissertation presents the challenges and the solutions provided in such a computing infrastructure

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

  5. ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)

    CERN Document Server

    Lewis, M E

    2000-01-01

    The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

  6. ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit

  7. Proceedings of International Workshop on Exception Handling in Object Oriented Systems: Towards Emerging Application Areas and New Programming Paradigms Workshop

    OpenAIRE

    Romanovsky, Alexander; Dony, Christophe; Knudsen, Jorgen Lindskov; Tripathi, Anand

    2003-01-01

    There are two trends in the development of modern object oriented systems: they are getting more complex and they have to cope with an increasing number of exceptional situations. The most general way of dealing with these problems is by employing exception handling techniques. Many object oriented mechanisms for handling exceptions have been proposed but there still are serious problems in applying them in practice. These are caused by:- complexity of exception code design and analysis,- not...

  8. Methodology on Investigating the Influences of Automated Material Handling System in Automotive Assembly Process

    Science.gov (United States)

    Saffar, Seha; Azni Jafar, Fairul; Jamaludin, Zamberi

    2016-02-01

    A case study was selected as a method to collect data in actual industry situation. The study aimed to assess the influences of automated material handling system in automotive industry by proposing a new design of integration system through simulation, and analyze the significant effect and influence of the system. The method approach tool will be CAD Software (Delmia & Quest). The process of preliminary data gathering in phase 1 will collect all data related from actual industry situation. It is expected to produce a guideline and limitation in designing a new integration system later. In phase 2, an idea or concept of design will be done by using 10 principles of design consideration for manufacturing. A full factorial design will be used as design of experiment in order to analyze the performance measured of the integration system with the current system in case study. From the result of the experiment, an ANOVA analysis will be done to study the performance measured. Thus, it is expected that influences can be seen from the improvement made in the system.

  9. Critical Issues for Long-Term Nuclear Waste Canister Safety: How 'Good' is 'Good Enough?'

    International Nuclear Information System (INIS)

    . The determination of an 'acceptable' impurity level for the canister material and an 'acceptable' flaw size (and shape) for any nondestructive evaluation (NDE) method employed to inspect the canisters may play a critical role in the prediction of long-term performance. Whether the lifetime of the canister is limited by the long-term creep rate of copper or the corrosion caused by the chemical composition of the water that re-saturates the Bentonite, the ultimate performance of the canister will be determined by a combination of degradation mechanisms for the copper barrier and the cast iron insert. Mechanical loadings will result in the deformation of the copper due to long-term creep or seismic shear. The corrosive effects of potential oxygen ingress during glaciation cycles, or due to the effects of microbiologically influenced corrosion (MIC), will result in the ultimate breaching of the canisters. Other potential corrosive effects include galvanic coupling between the copper and cast iron insert and the potential for radiolysis product formation shortly after emplacement. All of these mechanisms must be evaluated to determine whether the mechanical integrity of the KBS-3 canisters will be maintained for a sufficient time period to protect the health and safety of future generations. Performance assessment is the tool employed to evaluate the ability of the entire repository system to meet the regulatory requirements for safe, long-term performance. The mechanical integrity of the KBS-3 canister will significantly influence the results of all PA evaluations. Performance assessment studies are completed to evaluate the 'normal' or 'expected' degradation of the canisters. In addition, performance assessment models can be used to evaluate possible conditions that may result in accelerated canister degradation and failure. In reality, canisters will ultimately fail due to a combination of mechanisms and conditions and will exhibit a 'range' of lifetimes. The extent of

  10. A comparative life cycle assessment of material handling systems for sustainable mining.

    Science.gov (United States)

    Erkayaoğlu, M; Demirel, N

    2016-06-01

    In this comprehensive LCA comparison study, main objectives are to investigate life cycle environmental impacts of off-highway mining trucks and belt conveyors in surface mining. The research methodology essentially entails determination of the functional unit as 20,000 tons/day coal production transported for 5 km distance. After the system boundary was selected as the entire life cycle of material handling systems including pre-manufacturing of steel parts and plastic components, manufacturing, transportation, and utilization data was compiled from equipment manufacturers and the Eco-invent database. Life cycle impact categories for both material-handling systems were identified and the developed model was implemented using SIMAPRO 7.3. Climate change and acidification were selected as major impact categories as they were considered to be major concerns in mining industry. Although manufacturing stage had a significant impact on all of the environmental parameters, utilization stage was the hotspot for the selected impact categories. The results of this study revealed that belt conveyors have a greater environmental burden in climate change impact category when compared to the trucks. On the other hand, trucks have a greater environmental burden in acidification impact category when compared to the belt conveyors. This study implied that technological improvement in fuel combustion and electricity generation is crucial for the improvement of environmental profiles of off-highway trucks and belt conveyors in the mining industry. The main novelty of this study is that it is the first initiative in applying LCA in the Turkish mining industry. PMID:26986638

  11. Canister Storage Building (CSB) Hazard Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    POWERS, T.B.

    2000-03-16

    This report describes the methodology used in conducting the Canister Storage Building (CSB) Hazard Analysis to support the final CSB Safety Analysis Report and documents the results. This report describes the methodology used in conducting the Canister Storage Building (CSB) hazard analysis to support the CSB final safety analysis report (FSAR) and documents the results. The hazard analysis process identified hazardous conditions and material-at-risk, determined causes for potential accidents, identified preventive and mitigative features, and qualitatively estimated the frequencies and consequences of specific occurrences. The hazard analysis was performed by a team of cognizant CSB operations and design personnel, safety analysts familiar with the CSB, and technical experts in specialty areas. The material included in this report documents the final state of a nearly two-year long process. Attachment A provides two lists of hazard analysis team members and describes the background and experience of each. The first list is a complete list of the hazard analysis team members that have been involved over the two-year long process. The second list is a subset of the first list and consists of those hazard analysis team members that reviewed and agreed to the final hazard analysis documentation. The material included in this report documents the final state of a nearly two-year long process involving formal facilitated group sessions and independent hazard and accident analysis work. The hazard analysis process led to the selection of candidate accidents for further quantitative analysis. New information relative to the hazards, discovered during the accident analysis, was incorporated into the hazard analysis data in order to compile a complete profile of facility hazards. Through this process, the results of the hazard and accident analyses led directly to the identification of safety structures, systems, and components, technical safety requirements, and other

  12. Canister Storage Building (CSB) Hazard Analysis Report

    International Nuclear Information System (INIS)

    This report describes the methodology used in conducting the Canister Storage Building (CSB) Hazard Analysis to support the final CSB Safety Analysis Report and documents the results. This report describes the methodology used in conducting the Canister Storage Building (CSB) hazard analysis to support the CSB final safety analysis report (FSAR) and documents the results. The hazard analysis process identified hazardous conditions and material-at-risk, determined causes for potential accidents, identified preventive and mitigative features, and qualitatively estimated the frequencies and consequences of specific occurrences. The hazard analysis was performed by a team of cognizant CSB operations and design personnel, safety analysts familiar with the CSB, and technical experts in specialty areas. The material included in this report documents the final state of a nearly two-year long process. Attachment A provides two lists of hazard analysis team members and describes the background and experience of each. The first list is a complete list of the hazard analysis team members that have been involved over the two-year long process. The second list is a subset of the first list and consists of those hazard analysis team members that reviewed and agreed to the final hazard analysis documentation. The material included in this report documents the final state of a nearly two-year long process involving formal facilitated group sessions and independent hazard and accident analysis work. The hazard analysis process led to the selection of candidate accidents for further quantitative analysis. New information relative to the hazards, discovered during the accident analysis, was incorporated into the hazard analysis data in order to compile a complete profile of facility hazards. Through this process, the results of the hazard and accident analyses led directly to the identification of safety structures, systems, and components, technical safety requirements, and other

  13. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted

  14. The crane handling system for 500 litre drums of cemented radioactive waste

    International Nuclear Information System (INIS)

    As part of the AEA Technology strategy for dealing with radioactive wastes new waste treatment facilities are being built at the Winfrith Technology Centre (WTC), Dorset. One of the facilities at WTC is the Treated Radwaste Store (TRS) which is designed to store sealed 500 litre capacity drums of treated waste for an interim period until the national disposal facility is operational. Within the TRS two cranes have been incorporated, one spanning the entire width and travelling the length of the Store. The second operates within the area designated for drum handling during inspection work. The development of the design of these cranes and their associated control systems, to meet the complex requirements of operations whilst also satisfying the reliability and safety criteria, is discussed within the paper. (author)

  15. Film/chemistry selection for the earth resources technology satellite /ERTS/ ground data handling system

    Science.gov (United States)

    Shaffer, R. M.

    1973-01-01

    A detailed description is given of the methods of choose the duplication film and chemistry currently used in the NASA-ERTS Ground Data Handling System. The major ERTS photographic duplication goals are given as background information to justify the specifications for the desirable film/chemistry combination. Once these specifications were defined, a quantitative evaluation program was designed and implemented to determine if any recommended combinations could meet the ERTS laboratory specifications. The specifications include tone reproduction, granularity, MTF and cosmetic effects. A complete description of the techniques used to measure the test response variables is given. It is anticipated that similar quantitative techniques could be used on other programs to determine the optimum film/chemistry consistent with the engineering goals of the program.

  16. INfluence of vinasse on water movement in soil, using automatic acquisition and handling data system

    International Nuclear Information System (INIS)

    The vinasse, by-product of ethylic alcohol industry from the sugar cane juice or molasses yeast fermentation, has been incorporated in the soil as fertilizer, due to the its hight organic matter (2-6%), potassium and sulphate (0,1-0,5%) and other nutrient contents. By employing monoenergetic gamma-ray beam attenuation technique (241Am; 59,5 keV; 100 mCi) the influence of vinasse on the water movement in the soil was studied. For this, an automatic acquisition and handling data system was used, based in multichannel analyser, multi-scaling mode operated, coupled to a personal microcomputer and plotter. Despite the small depth studied (6 cm), it was observed that vinasse decreases the water infiltration velocity in the soil. (Author)

  17. The range of options for handling plane angle and solid angle within a system of units

    Science.gov (United States)

    Quincey, Paul

    2016-04-01

    The radian and steradian are unusual units within the SI, originally belonging to their own category of ‘supplementary units’, with this status being changed to dimensionless ‘derived units’ in 1995. Recent papers have suggested that angles could be handled in two different ways within the SI, both differing from the present system. The purpose of this paper is to provide a framework for putting such suggestions into context, outlining the range of options that is available, together with the advantages and disadvantages of these options. Although less rigorously logical than some alternatives, the present SI approach is generally supported, but with some changes to the SI brochure to make the position clearer, in particular with regard to the designation of the radian and steradian as derived units.

  18. Cantilever-based bio-chemical sensor integrated in a microliquid handling system

    DEFF Research Database (Denmark)

    Thaysen, Jacob; Marie, Rodolphe; Boisen, Anja

    2001-01-01

    The cantilevers have integrated piezoresistive readout which, compared to optical readout, enables simple measurements on even non-transparent liquids, such as blood. First, we introduce a simple theory for using piezoresistive cantilevers as surface stress sensors. Then, the sensor fabrication...... based on conventional microfabrication is described and the sensor characterization is discussed. During the characterization we found a stress sensitivity of (ΔR/R)=4.6:10 -4 (N/m)-1 and a minimum detectable surface stress change of 2.6 mN/m. Aqua regia etch of gold on top of the cantilevers has been...... monitored, and immobilization of single-stranded thiol modified DNA-oligos has been detected by the sensor. Finally, it is demonstrated that it is possible to analyze two samples simultaneously by utilizing the laminar flow in the microliquid handling system....

  19. Transporting existing VSC-24 canisters using a risk-based licensing approach

    International Nuclear Information System (INIS)

    The eventual disposition of the spent fuel assemblies loaded in canisters and casks currently designed and licensed only for on-site storage is an industry-wide issue. The canister-specific BUC evaluation approach developed by BFS can be used to license many of these storage canisters and casks for transportation. This will allow these storage canisters and casks to be transported intact to a long-term storage facility or repository, thereby minimizing fuel handling operations, impact on plant operations, and occupational exposure, as well as total infrastructure costs. Application of the proposed canister-specific BUC analysis approach to a preliminary evaluation of the 58 loaded MSBs demonstrates the benefits of this approach. The results of this preliminary evaluation show that a more rigorous analysis based on the known characteristics of the loaded spent fuel, rather than the design-basis fuel parameters, produces significantly lower maximum keff values and can be used to qualify many of the existing loaded storage canisters for transportation. Transportation certification for storage canisters having more reactive spent fuel payloads may require reliance on BUC approaches that are more aggressive than current NRC guidelines allow. Credit may be required for fission- product isotopes that do not have sufficient chemical assay data for benchmarking. In addition, reduced criticality safety margins may be required. For these more-aggressive BUC approaches, a risk assessment should be provided to support the NRC-approval basis. The risk assessment should evaluate the possibility and consequences of an accidental criticality event based upon inaccuracies in the characterization of the spent-fuel payloads

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. LHCB: Non-POSIX File System for the LHCB Online Event Handling

    CERN Multimedia

    Garnier, J-C; Cherukuwada, S S

    2010-01-01

    LHCb aims to use its O(20000) CPU cores in the High Level Trigger (HLT) and its 120 TB Online storage system for data reprocessing during LHC shutdown periods. These periods can last between a few days and several weeks during the winter shutdown or even only a few hours during beam interfill gaps. These jobs run on files which are staged in from tape storage to the local storage buffer. The result are again one or more files. Efficient file writing and reading is essential for the performance of the system. Rather than using a traditional shared filesystem such as NFS or CIFS we have implemented a custom, light-weight, non-Posix file-system for the handling of these files. Streaming this filesystem for the data-access allows to obtain high performance, while at the same time keep the resource consumption low and add nice features not found in NFS such as high-availability, transparent failover of the read and write service. The writing part of this file-system is in successful use for the Online, real-time w...

  4. Improving Vehicle Ride and Handling Using LQG CNF Fusion Control Strategy for an Active Antiroll Bar System

    OpenAIRE

    Zulkarnain, N.; H. Zamzuri; Y. M. Sam; Mazlan, S. A.; S. M. H. F. Zainal

    2014-01-01

    This paper analyses a comparison of performance for an active antiroll bar (ARB) system using two types of control strategy. First of all, the LQG control strategy is investigated and then a novel LQG CNF fusion control method is developed to improve the performances on vehicle ride and handling for an active antiroll bar system. However, the ARB system has to balance the trade-off between ride and handling performance, where the CNF consists of a linear feedback law and a nonlinear feedback ...

  5. Canister storage building natural phenomena design loads

    International Nuclear Information System (INIS)

    This document presents natural phenomena hazard (NPH) loads for use in the design and construction of the Canister Storage Building (CSB), which will be located in the 200 East Area of the Hanford Site

  6. Canister transfer into repository in shaft alternative

    International Nuclear Information System (INIS)

    In this report, a study of lift transportation of a massive canister for spent nuclear fuel is considered. The canister is transferred from ground level to repository, which lies in the depth of 400 to 500 m in the bedrock. The canister is a massive metal vessel, whose weight is 19 to 29 tons, and which is strongly irradiant (gamma and neutrons), and which contains 1.4 to 2.2 tons of very strongly radio-active material, the activity of the fuel should not be spread in the environment even during postulated accidents. The study observes that the lift alternative is possible to be built and through good design practices and good maintenance procedures its safety, reliability and usability can be kept on such high level that canister transport is estimated to be licensable. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  8. Force Sensitive Handles and Capacitive Touch Sensor for Driving a Flexible Haptic-Based Immersive System

    Directory of Open Access Journals (Sweden)

    Umberto Cugini

    2013-10-01

    Full Text Available In this article, we present an approach that uses both two force sensitive handles (FSH and a flexible capacitive touch sensor (FCTS to drive a haptic-based immersive system. The immersive system has been developed as part of a multimodal interface for product design. The haptic interface consists of a strip that can be used by product designers to evaluate the quality of a 3D virtual shape by using touch, vision and hearing and, also, to interactively change the shape of the virtual object. Specifically, the user interacts with the FSH to move the virtual object and to appropriately position the haptic interface for retrieving the six degrees of freedom required for both manipulation and modification modalities. The FCTS allows the system to track the movement and position of the user’s fingers on the strip, which is used for rendering visual and sound feedback. Two evaluation experiments are described, which involve both the evaluation and the modification of a 3D shape. Results show that the use of the haptic strip for the evaluation of aesthetic shapes is effective and supports product designers in the appreciation of the aesthetic qualities of the shape.

  9. Evolution of radiometric instrumentation systems on operating reprocessing and waste handling facilities

    International Nuclear Information System (INIS)

    A wide variety of special purpose radiometric instrumentation systems are now in use in the Sellafield Magnox fuel reprocessing facility and in the associated plants and waste handling operations. These systems, which were developed by the Physical Science and Engineering Development Group have accumulated many years of successful operation. However as processes have been improved and regulatory requirements have become increasingly stringent, it has become necessary to re-assess the monitoring approach employed in a number of these devices. As examples of this process of evolution, which is being carried out in parallel with the development of other special instrumentation for the new plants now under construction at Sellafield, the paper outlines the way in which two of the existing devices are currently being replaced by more sophisticated and versatile systems. For the monitoring of crated waste, a flexible neutron coincidence monitor will incorporate high resolution gamma spectrometry to provide a more reliable overall assessment of plutonium and to assess the presence and distribution of U-235 contamination. Swarf arising from the decanning of Magnox fuel will be monitored by high resolution gamma spectrometry and the data analysis extended in order to provide a quantitative assessment of some 40 specified radioisotopes

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

    International Nuclear Information System (INIS)

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

  11. Pressure tests of two KBS-3 canister mock-ups

    International Nuclear Information System (INIS)

    The Swedish concept for geological disposal of spent nuclear fuel, the so-called KBS-3 concept, relies on a multibarrier system with the copper/cast iron canister as the first barrier. The canister is designed to retain its integrity for at least 100,000 years, which means that future glaciations need to be considered. A 3 km thick ice block together with hydrostatic pressure from groundwater and swelling of the buffer material would produce hydrostatic compressive stresses of maximum 44 MPa (440 bar). Although the canister is loaded globally in compression, tensile stresses develop at fuel channel surface with increasing load. Tensile tests of the insert material in the development phase of the KBS-3 canister indicated a large scatter and relatively low values of the inserts' ductility. An important issue was whether this could lead to mechanical failure of canisters at the 44 MPa iso-static load either by plastic collapse or fracture from the defects in the regions with tensile stresses. SKB therefore initiated a project together with the European commission's Joint Research Centre (JRC) Institute of Energy in Petten and a number of Swedish partners to evaluate the probability of mechanical failure during glaciation. Three inserts manufactured by different Swedish foundries and referred to as 1, 125 and 126 were used in the project. A large statistical test programme was developed to determine statistical distributions of various material parameters and defect distributions. These data were subsequently used in probabilistic analysis to determine the probability for local plastic collapse or fracture. The main conclusion was that the failure probability is extremely low at the design load (44 MPa) provided some basic geometrical requirements are fulfilled. In parallel to the statistical test programme and the associated analysis, the group decided also to perform two pressure tests of canister mock-ups to demonstrate the actual safety margins. The fractographic

  12. Biological Research in Canisters (BRIC) - Light Emitting Diode (LED)

    Science.gov (United States)

    Levine, Howard G.; Caron, Allison

    2016-01-01

    The Biological Research in Canisters - LED (BRIC-LED) is a biological research system that is being designed to complement the capabilities of the existing BRIC-Petri Dish Fixation Unit (PDFU) for the Space Life and Physical Sciences (SLPS) Program. A diverse range of organisms can be supported, including plant seedlings, callus cultures, Caenorhabditis elegans, microbes, and others. In the event of a launch scrub, the entire assembly can be replaced with an identical back-up unit containing freshly loaded specimens.

  13. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    International Nuclear Information System (INIS)

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  14. Statistical analysis of DWPF reference canister dimensions

    International Nuclear Information System (INIS)

    Twenty dimensional measurements were conducted on seven empty Defense Waste Processing Facility (DWPF) reference canisters. These measurements were repeated after the canisters were filled with simulated nuclear waste glass. An in-depth statistical analysis of the results indicated that changes do occur as a result of filling the steel canisters with glass poured at 1150 degree C for four of the parameters. While small, these changes were statistically significant. The analysis indicates the maximum dimensional change found to occur after the filling for each variable. Statistical tests were used to determine if canister dimensions do significantly change, and corresponding variance information is presented. The results showed that the four measured parameters affected by filling are bottom diameter, bottom end diameter flange tilt, and lower head mismatch. Significant variability also existed for height, upper weld, ID label, lower head mismatch, and lower head ovality due to the measurements coming from different canisters. Finally, lower head mismatch showed variability caused by the data being taken at different locations on the canister. This location effect did not affect any of the other variables in this way

  15. Am/Cm canister temperature evaluation in CIM5

    International Nuclear Information System (INIS)

    To facilitate the evaluation of alternate canister designs, 2 canisters were outfitted with thermocouples at elevations of 1/2, 3 1/2, and 6 1/2 inches from the canister bottom. The canisters were fabricated from two inch diameter schedule 10 and two inch diameter schedule 40 stainless steel pipe. Each canister was filled with approximately 2 kilograms of 49 wt percent lanthanide (Ln) loaded 25SrABS glass during 5 inch Cylindrical Induction Melter (CIM5) runs for TTR Tasks 3.03 and 4.03. Melter temperature, total mass of glass poured, and the glass pour rates were almost identical in both runs. The schedule 40 canister has a slightly smaller ID compared to the schedule 10 canister and therefore filled to a level of 9.5 inches compared to 8.0 inches for the schedule 40 canister. The schedule 40 canister had an empty mass of 1906 grams compared to 919 grams for the schedule 10 canister. The schedule 10 canister was found to have a higher maximum surface temperature by about 50--100 C (depending on height) during the glass pour compared to the schedule 40 canister. The additional thermal mass of the schedule 40 canister accounts for this difference. Once filled with glass, each of the canisters cooled at about the same rate, taking about an hour to cool below a maximum surface temperature of 200 C. No significant deformation of the either of the canisters was visually observed

  16. Inspection of copper canisters for spent nuclear fuel by means of Ultrasonic Array System. Electron beam evaluation, modeling and materials characterization

    International Nuclear Information System (INIS)

    Research conducted in the fifth phase of the SKB's study aimed at developing ultrasonic techniques for assessing EB welds copper canisters is reported here. This report covers three main tasks: evaluation of electron beam (EB) welds, modeling of ultrasonic fields and characterization of copper material. A systematic analysis of ultrasonic interaction and imaging of an EB weld has been performed. From the analysis of histograms of the weld ultrasonic image, it appeared that the porosity tended to be concentrated towards the upper side of a HV weld, and a guideline on how to select the gates for creating C-scans has been proposed. The spatial diversity method (SDM) has shown a limited ability to suppress grain noise both in the parent material (copper) and in the weld so that the ultrasonic image of the weld could be improved. The suppression was achieved at the price of reduced spatial resolution. The ability of wavelet filters to enhance flaw responses has been studied. An FIR (finite impulse response) filter, based on Sombrero mother wavelet, has yield encouraging results concerning clutter suppression. However, the physical explanation for the results is still missing and needs further research. For modeling of ultrasonic fields of the ALLIN array, an approach to computing the SIR (spatial impulse response) of a cylindrically curved, rectangular aperture has been developed. The aperture is split into very narrow strips in the cylindrically curved direction and SIR of the whole aperture by superposing the individual impulse responses of those strips. Using this approach, the SIR of the ALLIN array with a cylindrically curved surface has been calculated. The pulse excitation of normal velocity on the surface of the array, that is required for simulating actual ultrasonic fields, has been determined by measurement in combination with a deconvolution technique. Using the SIR and the pulse excitation obtained, the pulsed-echo fields from the array have been simulated

  17. Inspection of copper canisters for spent nuclear fuel by means of Ultrasonic Array System. Electron beam evaluation, modeling and materials characterization

    Energy Technology Data Exchange (ETDEWEB)

    Ping Wu; Lingvall, F.; Stepinski, T. [Uppsala Univ. (Sweden). Dept. of Material Science

    1999-12-01

    Research conducted in the fifth phase of the SKB's study aimed at developing ultrasonic techniques for assessing EB welds copper canisters is reported here. This report covers three main tasks: evaluation of electron beam (EB) welds, modeling of ultrasonic fields and characterization of copper material. A systematic analysis of ultrasonic interaction and imaging of an EB weld has been performed. From the analysis of histograms of the weld ultrasonic image, it appeared that the porosity tended to be concentrated towards the upper side of a HV weld, and a guideline on how to select the gates for creating C-scans has been proposed. The spatial diversity method (SDM) has shown a limited ability to suppress grain noise both in the parent material (copper) and in the weld so that the ultrasonic image of the weld could be improved. The suppression was achieved at the price of reduced spatial resolution. The ability of wavelet filters to enhance flaw responses has been studied. An FIR (finite impulse response) filter, based on Sombrero mother wavelet, has yield encouraging results concerning clutter suppression. However, the physical explanation for the results is still missing and needs further research. For modeling of ultrasonic fields of the ALLIN array, an approach to computing the SIR (spatial impulse response) of a cylindrically curved, rectangular aperture has been developed. The aperture is split into very narrow strips in the cylindrically curved direction and SIR of the whole aperture by superposing the individual impulse responses of those strips. Using this approach, the SIR of the ALLIN array with a cylindrically curved surface has been calculated. The pulse excitation of normal velocity on the surface of the array, that is required for simulating actual ultrasonic fields, has been determined by measurement in combination with a deconvolution technique. Using the SIR and the pulse excitation obtained, the pulsed-echo fields from the array have been

  18. EEVADO-ontology based construction of a knowledge based system for handling radiation emergency

    International Nuclear Information System (INIS)

    The basic aim of handling a radiation emergency is to protect the site personnel and public in the off-site areas. Hence it is required to assess the accident to develop the protective action recommendations for the site personnel, members of the public and emergency workers. An explanation facility is required to analyze the emergency situation and make needful decision. Algorithmic based software cannot provide this support as it does not have any such facility. A knowledge based system (KBS) obviously would act as a special purpose intelligent agent on behalf or at the behest of the emergency organization. Knowledge of the KBS has been elicited on the basis of ontology, a set of definitions of classes, objects, attributes, relations and constraints. Ontology provides a vocabulary for the expression of domain knowledge. The present paper discusses the development and utility of a knowledge based system - EEVADO (Emergency Evaluation and Dosimetry). The present software is verified and validated by ontology and by conducting Desktop Exercise in training program on Emergency Preparedness. (author)

  19. The Conceptual Design of a Mechatronic System to Handle Bedridden Elderly Individuals

    Directory of Open Access Journals (Sweden)

    Silva Bruno

    2016-05-01

    Full Text Available The ever-growing percentage of elderly people in developed countries have made Ambient Assisted Living (AAL solutions an important subject to be explored and developed. The increase in geriatric care requests are overburdening specialized institutions that cannot cope with the demand for support. Patients are forced to have to remain at their homes encumbering the spouse or close family members with the caregiver role. This caregiver is not always physically and technically apt to assist the bedridden person with his/her meals and hygiene/bath routine. Consequently, a solution to assist caregivers in these tasks is of the utmost importance. This paper presents an approach for supporting caregivers when moving and repositioning Bedridden Elderly Peoples (BEPs in home settings by means of a mechatronic system inspired by industrial conveyers. The proposed solution is able to insert itself underneath the patient, due to its low-profile structural properties, and retrieve and reallocate him/her. Ideally, the proposed mechatronic system aims to promote autonomy by reducing handling complexity, alter the role of the caregiver from physically handler of the BEP to an operator/supervisor role, and lessen the amount of effort expended by caregivers and BEPs alike.

  20. The Conceptual Design of a Mechatronic System to Handle Bedridden Elderly Individuals.

    Science.gov (United States)

    Bruno, Silva; José, Machado; Filomena, Soares; Vítor, Carvalho; Demétrio, Matos; Karolina, Bezerra

    2016-01-01

    The ever-growing percentage of elderly people in developed countries have made Ambient Assisted Living (AAL) solutions an important subject to be explored and developed. The increase in geriatric care requests are overburdening specialized institutions that cannot cope with the demand for support. Patients are forced to have to remain at their homes encumbering the spouse or close family members with the caregiver role. This caregiver is not always physically and technically apt to assist the bedridden person with his/her meals and hygiene/bath routine. Consequently, a solution to assist caregivers in these tasks is of the utmost importance. This paper presents an approach for supporting caregivers when moving and repositioning Bedridden Elderly Peoples (BEPs) in home settings by means of a mechatronic system inspired by industrial conveyers. The proposed solution is able to insert itself underneath the patient, due to its low-profile structural properties, and retrieve and reallocate him/her. Ideally, the proposed mechatronic system aims to promote autonomy by reducing handling complexity, alter the role of the caregiver from physically handler of the BEP to an operator/supervisor role, and lessen the amount of effort expended by caregivers and BEPs alike. PMID:27213383

  1. The Conceptual Design of a Mechatronic System to Handle Bedridden Elderly Individuals

    Science.gov (United States)

    Bruno, Silva; José, Machado; Filomena, Soares; Vítor, Carvalho; Demétrio, Matos; Karolina, Bezerra

    2016-01-01

    The ever-growing percentage of elderly people in developed countries have made Ambient Assisted Living (AAL) solutions an important subject to be explored and developed. The increase in geriatric care requests are overburdening specialized institutions that cannot cope with the demand for support. Patients are forced to have to remain at their homes encumbering the spouse or close family members with the caregiver role. This caregiver is not always physically and technically apt to assist the bedridden person with his/her meals and hygiene/bath routine. Consequently, a solution to assist caregivers in these tasks is of the utmost importance. This paper presents an approach for supporting caregivers when moving and repositioning Bedridden Elderly Peoples (BEPs) in home settings by means of a mechatronic system inspired by industrial conveyers. The proposed solution is able to insert itself underneath the patient, due to its low-profile structural properties, and retrieve and reallocate him/her. Ideally, the proposed mechatronic system aims to promote autonomy by reducing handling complexity, alter the role of the caregiver from physically handler of the BEP to an operator/supervisor role, and lessen the amount of effort expended by caregivers and BEPs alike. PMID:27213383

  2. An assessment of canister needs for defueling the TMI-2 core

    International Nuclear Information System (INIS)

    It is projected that the TMI-2 Cleanup Program can be completed with a total of 355 canisters (272 fuel, 75 filter, and 8 k/o canisters). This is within the 360 canister space allocation at the INEL. There is a sufficient number and mix of available canisters on-site to meet the outstanding requirements. As of May 1989, the shipment campaign has included 18 rail shipments, with a total of 259 canisters. It is estimated that an additional five rail shipments of three casks (21 canisters) each will be required to complete the program. The achievements of the shipment campaign, the challenges that have been presented, and the reasons for its success can be outlined as follows: very few reactors have ever had to undertake a fuel shipment program paralleled to the magnitude of the TMI-2 program; the cleanup project faced a task of transporting an entire damaged reactor core from TMI-2 to the INEL; this shipment campaign may one day become a blueprint for future shipments of spent fuel by other utilities; the transport system essentially consists of three major subsystems: the casks, the cask support systems, and the shielded dry fuel transfer system, the program successfully worked out the interactions and operation of these subsystems; to date, the shipment program has compiled an impressive record of safe, on-time, and essentially trouble-free performance

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

    International Nuclear Information System (INIS)

    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 D20 hydraulics, H20 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 have

  4. Handling technology of low decontaminated TRU fuel for the simplified pelletizing method fuel fabrication system

    International Nuclear Information System (INIS)

    glovebox type equipment, because equipment are installed in hot cells. In particular, it is essential to establish repairing system in the hot cell, because the equipment consist of machines required to have high precision and operators can not maintain these machines directly. We proposed basic concept for repairing and maintenance system consisted of three stages; (a) replace the out ordered module in the main process cell, (b) decontaminate and roughly disassemble the module in the maintenance cell and (c) refurbish the module using the globe box in the maintenance room. To confirm feasibility of this concept, representatives of in-cell equipment, a pressing machine, a pellet inspection equipment and some powder analysers have been investigated by cold mock-up examination. A pressing machine is favorable for testing feasibility of modularized equipment because the process uses various machining. A pellet inspection equipment and powder analysers are favorable for developing how to maintain such precise and sensitive equipment. Modularized pressing machine was designed conceptually consisting 30 modules with 70 kinds of maintenance terms. In order to assist operation on replacing modules, a robot arm manipulator is introduced. Also, a small size robot arm for handling a pellet is developing as well. Cold mock-up tests using modularized equipment and handling equipment will be completed by the end of JFY 2009. 2. Development of heat removal system. High heat generation by decay heat of MAs causes much undesirable effects to fuel quality, such as, degradation of organic additives, re-oxidation of fuel. Development of heat removal system is necessary for realizing mass-production plant, even though the simplified pelletizing process brings some advantage on this issue (We have adopted less inorganic additives process; binder-less granulation and die wall lubrication pressing process. This adoption can solve the problem of potential evaporation of the additives caused by

  5. Criticality Safety Evaluation Report for the Cold Vacuum Drying (CVD) Facility's Process Water Handling System

    International Nuclear Information System (INIS)

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified

  6. Criticality safety evaluation report for the cold vacuum drying facility's process water handling system

    International Nuclear Information System (INIS)

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified

  7. System simulation on fractionation radiation doses and radioisotope handling in Nuclear medicine

    International Nuclear Information System (INIS)

    This paper describes the practical and theoretical learning of students from Medical Physics course at the Fundacao Universidade Federal do Rio Grande (FURG) on fractionation radiation doses, radioisotope handling and elution of molybdenum generators (Mo-99) / technetium (Tc -99m)

  8. Criticality Safety Evaluation Report for the Cold Vacuum Drying (CVD) Facilities Process Water Handling System

    Energy Technology Data Exchange (ETDEWEB)

    KESSLER, S.F.

    2000-08-10

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified.

  9. Event Handling in the OpenModelica Compiler and Run-time System

    OpenAIRE

    Lundvall, Håkan; Fritzson, Peter

    2005-01-01

    The paper gives an introduction to the problem simulating hybrid DAEs with event-handling using the Modelica language. An implementation in the OpenModelica compiler is presented, and some preliminary results are reported.

  10. Examples of automated material handling systems for iron and steel works; Tekko denroyo ni okeru jidoka butsuryu system no jirei

    Energy Technology Data Exchange (ETDEWEB)

    Tamura, Y.; Sugimoto, T.; Kimura, M.; Kudo, M. [Hitachi, Ltd., Tokyo (Japan)

    1996-06-01

    The iron and steel industry in which reduction in production cost and improvement in quality management are the most important problem has positively introduced systems to identify material flows in a factory on a real-time basis and achieve power saving in material moving. This paper shows a sequence of introducing a material handling system proposed by Hitachi, Ltd., introduces examples of the actual installations, and describes the latest material handling automation technology. Examples of actual installation may include a case of automating a scrap yard. In the case, work of unloading materials delivered on trucks of unspecified shapes is separated from a work to move them into and from a warehouse in the yard, or a buffer hopper is installed to level out operation time of cranes. Thus, automation and on-line use of information were achieved upon changing the yard work procedure. For example, load weight on a hopper is taken in on a real-time basis to use the result for controlling the number of deliveries out from a warehouse, improving the mixing accuracy, and managing the inventory information. 6 refs., 4 figs.

  11. Process and machinery description of equipment for deposition of canisters in medium-long deposition holes

    International Nuclear Information System (INIS)

    In this report twelve methods are presented to deposit a canister with spent nuclear fuel in a horizontal hole, several canisters per hole (MLH). These methods are part of the KBS-3 system. They have been developed successively, after an analysis of weak points and strong points in previously described methods. In conformance with the guidelines for Project JADE, a choices of system has been considered during the development work. This is whether canister and bentonite buffer should be deposited 'in parts', i.e. at different occasions, but shortly after each other or 'in a package', i.e. together in a single package. The other choice in the guidelines for the JADE project, whether the canister should be placed in a radiation shield or not during transport in the secondary tunnels, was not relevant to MLR. The basic technical problem is depositing heavy objects, the canister and the buffer components, in an horizontal hole which is approximately 200 m deep. Two methods for depositing of the bentonite barrier and the canisters in separate processes have been studied. For depositing of the bentonite barrier and the canister 'in a package', four alternative techniques have been studied: a metallic sleeve around the package, a loading scoop that is rotated, a fork carriage and rails. The repeated transports in a hole, a consequence of depositing several canisters in the same hole, could lead to the rock being crushed. The mutual impact of machines, load and rock wall has therefore been particularly considered. In several methods, the use of a gangway has been proposed (steel plates or layer of ice). A failure mode and effect analysis has been performed for one of the twelve methods. When comparing with a method to deposit one canister per hole using the same technique, the need for equipment and resources is far larger for this MLH method if incidents should occur during depositing. The development work reported here has not yet yielded a definitive method for placing

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  14. Interpretation and handling manual of fusion experimental data monitor system DAISY version 1

    International Nuclear Information System (INIS)

    We have developed a new soft program 'DAISY' to monitor fusion experimental data on graphic terminals. Users of 'DAISY' are free from the troublesome work to access data base and to lean computer graphic language, and can interactively display their abjective data with easy keyboard handling. 'DAISY' provides many display patterns, quick control of display figures and easy handling of plot data. This report summarizes a brief interpretation and the usage of 'DAISY'. (author)

  15. The influence on biogas production of three slurry-handling systems in dairy farms

    Directory of Open Access Journals (Sweden)

    Damiano Coppolecchia

    2015-04-01

    Full Text Available Handling systems can influence the production of biogas and methane from dairy farm manures. A comparative work performed in three different Italian dairy farms showed how the most common techniques (scraper, slatted floor, flushing can change the characteristics of collected manure. Scraper appears to be the most neutral choice, as it does not significantly affect the original characteristics of manure. Slatted floor produces a manure that has a lower methane potential in comparison with scraper, due to: a lower content of volatile solids caused by the biodegradation occurring in the deep pit, and a lower specific biogas production caused by the change in the characteristics of organic matter. Flushing can produce three different fluxes: diluted flushed manure, solid separated manure and liquid separated manure. The diluted fraction appears to be unsuitable for conventional anaerobic digestion in completely stirred reactors (CSTR, since its content of organic matter is too low to be worthwhile. The liquid separated fraction could represent an interesting material, as it appears to accumulate the most biodegradable organic fraction, but not as primary substrate in CSTR as the organic matter concentration is too low. Finally, the solid-liquid separation process tends to accumulate inert matter in the solid separated fraction and, therefore, its specific methane production is low.

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

    International Nuclear Information System (INIS)

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

  17. Low cost, robust and real time system for detecting and tracking moving objects to automate cargo handling in port terminals

    NARCIS (Netherlands)

    Vaquero, V.; Repiso, E.; Sanfeliu, A.; Vissers, J.; Kwakkernaat, M.

    2016-01-01

    The presented paper addresses the problem of detecting and tracking moving objects for autonomous cargo handling in port terminals using a perception system which input data is a single layer laser scanner. A computationally low cost and robust Detection and Tracking Moving Objects (DATMO) algorithm

  18. RFNC-VNIIEF experience in development and operation of hydrogen isotopes gas-handling systems for basic research

    Energy Technology Data Exchange (ETDEWEB)

    Yukhimchuk, A. A.; Il' kaev, R. I. [Russian Federal Nuclear Center, All - Russian Research Inst. of Experimental Physics, Mira av., 37, Sarov, Nizhny Novgorod region, 607188 (Russian Federation)

    2008-07-15

    Application of hydrogen isotopes in different fields of fundamental physics obtained by RFNC-VNIIEF in the last decade are presented. Gas-handling systems for scientific experiments, some technologies and designs of the setup key elements are described, and results obtained with the developed equipment are outlined. (authors)

  19. 5-DOF Dynamic Model of Vehicle Shimmy System with Clearance at Universal Joint in Steering Handling Mechanism

    OpenAIRE

    Jianwei Lu; Yi Xu; Chen Hu; Vakakis, Alexander F.; Bergman, Lawrence A.

    2013-01-01

    5-DOF dynamic model of vehicle shimmy system with clearance in universal joint of steering handling mechanism is presented. The sub model of cross shaft universal joint with clearance is built based on Hertz' theory, and two-state model is applied to describe the contact force. The sub model of the universal joint is combined with the simplified dynamic model of steering system, and a 5-DOF dynamic model of vehicle shimmy system with consideration of assembling clearance in universal joint of...

  20. Procedural development for nuclear waste canister impact testing

    International Nuclear Information System (INIS)

    Double containment requirements for transporting nuclear waste in glass form are costly and may not be necessary for some waste forms. To allow single containment, a procedure for examining particle size distribution and the amount of respirable particles generated under accident conditions was needed. A statistically designed experiment was conducted to examine the effects of glass temperature, fill rate and canister drop orientation upon the amount of sub-ten micron particles generated under simulated accident conditions. Measuring such small particles is somewhat inaccurate because of material loss in handling. By assuming a lognormal particle size distribution, the amount of sub-ten micron particles was estimated from the results for the larger measurable particles. Analyses revealed no temperature or fill rate effect but indicated that the amount of respirable particles is affected by drop orientation. This led to identification of a worst case drop orientation to be used in qualification testing. 4 refs., 2 figs

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

    International Nuclear Information System (INIS)

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

  2. 5-DOF Dynamic Model of Vehicle Shimmy System with Clearance at Universal Joint in Steering Handling Mechanism

    Directory of Open Access Journals (Sweden)

    Jianwei Lu

    2013-01-01

    Full Text Available 5-DOF dynamic model of vehicle shimmy system with clearance in universal joint of steering handling mechanism is presented. The sub model of cross shaft universal joint with clearance is built based on Hertz' theory, and two-state model is applied to describe the contact force. The sub model of the universal joint is combined with the simplified dynamic model of steering system, and a 5-DOF dynamic model of vehicle shimmy system with consideration of assembling clearance in universal joint of steering handling mechanism is presented. Based on this model, numerical analysis is carried out to evaluate the influence of clearance in universal joint on the dynamic behavior of the vehicle shimmy system. The results show that the clearance and some other parameters, such as vehicle speed, have coupled contribution to the dynamic behavior of the vehicle shimmy system. The conclusions provide theoretical basis for effective attenuation of vehicle shimmy, especially for those in-service vehicles.

  3. Drop Testing Representative Multi-Canister Overpacks

    Energy Technology Data Exchange (ETDEWEB)

    Snow, Spencer D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morton, Dana K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    The objective of the work reported herein was to determine the ability of the Multi- Canister Overpack (MCO) canister design to maintain its containment boundary after an accidental drop event. Two test MCO canisters were assembled at Hanford, prepared for testing at the Idaho National Engineering and Environmental Laboratory (INEEL), drop tested at Sandia National Laboratories, and evaluated back at the INEEL. In addition to the actual testing efforts, finite element plastic analysis techniques were used to make both pre-test and post-test predictions of the test MCOs structural deformations. The completed effort has demonstrated that the canister design is capable of maintaining a 50 psig pressure boundary after drop testing. Based on helium leak testing methods, one test MCO was determined to have a leakage rate not greater than 1x10-5 std cc/sec (prior internal helium presence prevented a more rigorous test) and the remaining test MCO had a measured leakage rate less than 1x10-7 std cc/sec (i.e., a leaktight containment) after the drop test. The effort has also demonstrated the capability of finite element methods using plastic analysis techniques to accurately predict the structural deformations of canisters subjected to an accidental drop event.

  4. Evaluation of a handling system for ley crop used in biogas production. Capacities and costs for a centralised system

    Energy Technology Data Exchange (ETDEWEB)

    Vaagstroem, Lena

    2005-07-01

    Within the Vaextkraft project in Vaesteraas, Sweden, biogas is to be produced out of ley crop and organic waste. The aim of this study has been to estimate the capacities within the handling system used for the ley crop harvest, and the resources needed. For this purpose a model in the form of a calculation program in Excel was built. The model makes it possible to vary parameters such as transport system design, distance from fields to storage, dry matter content and yield. The results showed that it is essential to match the capacities between chopper and transport to minimise the time and cost connected to the harvest. To avoid creating costly bottlenecks in the Vaextkraft case the transport system has to consist of at least two trucks with trailers. The estimates made with the model suggests that the distance to storage is strongly linked to the dimensioning of the transport system, whereas the number of fields and their size has a lesser impact on harvest time and cost. Variation of the dry matter yield from a base scenario had an impact on the cost for harvesting, but not on the choice of transport system. The model couldn't detect any reliable differences in total costs due to the variation of dry matter content between 25-45%. The choice of chopping machinery in the Vaextkraft project leads to small timeliness costs. Together with a reasonable choice of transport system they will only constitute a few percent of the total costs for harvesting.

  5. Technical specifications of air handling system of the inspection of a field

    International Nuclear Information System (INIS)

    A T.A.B(Testing, Adjusting and Balancing) technique, the basic technique of air handling facility, is one of the essential technical items which workers in charge of operation of facilities have to acquire. Especially, through scientific and reasonable inspective procedures, the reduction of energy and guarantee of designed skill have become influential important problems in our time rather than in the past days. Entrepreneurs have required more thorough verify of performances and procedure of test in order to raise the investment efficiency and reduce expenditure. For that reason, I hope that co-operator acquire objective and substantial knowledges about air handling facility so that they are helped from them

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

    International Nuclear Information System (INIS)

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

  7. Easy-handling spectrum-management on-line system MMCU for the μ-VAX based data acquisition system in the SF cyclotron facility

    International Nuclear Information System (INIS)

    An easy-handling spectrum-management system for both on-line and off-line analysis has been developed. This system 'Man Machine Communication Utility (MMCU)' is aimed to unify the data acquisition system of INS, CYRIC (Tohoku Univ.) and RIKEN. This system has been successfully tested in a nuclear astrophysical experiment at INS. (author)

  8. Canister displacement in KBS-3V. A theoretical study

    International Nuclear Information System (INIS)

    The vertical displacement of the canister in the KBS-3V concept has been studied in a number of consolidation and creep calculations using the FE-program ABAQUS. The creep model used for the calculations is based on Singh-Mitchell's creep theory, which has been adapted to and verified for the buffer material MX-80 in earlier tests. A porous elastic model with Drucker-Prager plasticity has been used for the consolidation calculations. For simplicity the buffer has been assumed to be water saturated from start. In one set of calculations only the consolidation and creep in the buffer without considering the interaction with the backfill was studied. In the other set of calculations the interaction with the backfill was included for a backfill consisting of an in situ compacted mixture of 30% bentonite and 70% crushed rock. The motivation to also study the behaviour of the buffer alone was that the final choice of backfill material and backfilling technique is not made yet so that set of calculations simulates a backfill that has identical properties with the buffer. The two cases represent two extreme cases, one with a backfill that has a low stiffness and the lowest allowable swelling pressure and one that has the highest possible swelling pressure and stiffness. The base cases in the calculations correspond to the final average density at saturation of 2,000 kg/m3 with the expected swelling pressure of 7 MPa in a buffer. In order to study the sensitivity of the system to loss in bentonite mass and swelling pressure seven additional calculations were done with reduced swelling pressure down to 80 kPa corresponding to a density at water saturation of about 1,500 kg/m3. The calculations included two stages, where the first stage models the swelling and consolidation that takes place in order for the buffer to reach force equilibrium. This stage takes place during the saturation phase and the subsequent consolidation/swelling phase. The second stage models the

  9. Development of a flexible dialysis pore water sampler placement system: easy handling and related error sources

    Directory of Open Access Journals (Sweden)

    Stephan Hilgert

    2014-02-01

    Full Text Available Investigations in the context of greenhouse gas production measurements in sub-tropical reservoirs brought up the necessity to survey the in situ pore water gas and ion concentrations at many positions within a relatively short time. As several sediment cores were taken, the interest in analyzing the pore water at the same time and at the same positions forced us to develop a cost- and time saving method for the placement of dialysis pore water samplers (DPS. General prerequisites were the ability to place several DPS per day, within a flexible depth range of up to 40 m and with a low cost budget. To meet these requirements, a DPS placing system (DPSPS was developed, which would allow the precise placement of DPS in water with a depth of up to 40 m and assessing the biases of on-board measurements and possible methodological improvements. The DPSPS was transported to Brazil and tested in a measurement campaign for 10 days. The measurements were carried out during two campaigns in December 2012 and March 2013 in the Capivari Reservoir north-east of Curitiba in the State of Paraná. The system worked properly and several DPS could be placed from a 5 m class aluminum boat. The placement was performed with high accuracy regarding the positioning as well as the penetration depth of the DPS. After the recovery of the DPS, the possible biases during sampling were analyzed. Possible back-diffusion was investigated, taking oxygen concentration as one representative parameter for estimation of the sample behavior. Laboratory as well as field results showed that special care has to be taken to minimize the influence of diffusion processes during post-recovery sampling. The results also suggested that the used membranes are affected by clogging which is likely to influence the diffusion times of various ions and gases. It can be stated that the DPSPS was developed successfully as the demands in terms of handling as well as monitoring efficiency and sample

  10. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a join ted arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  11. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  12. Interim Storage of RH-TRU 72B Canisters at the DOE Oak Ridge Reservation

    International Nuclear Information System (INIS)

    This paper describes an evaluation performed by the Department of Energy (DOE) Oak Ridge Operations (ORO) office for potential interim storage of remote-handled (RH) transuranic (TRU) 72B waste canisters at the Oak Ridge National Laboratory (ORNL). The evaluation included the conceptual design of a devoted canister storage facility and an assessment of the existing RHTRU waste storage facilities for storage of canisters. The concept for the devoted facility used modular concrete silos located on an above-grade storage pad. The assessment of the existing facilities considered the potential methods, facility modifications, and conceptual equipment that might be used for storage of 400 millisievert per hour (mSv/hr) canisters. The results of the evaluation indicated that the initial investment into a devoted facility was relatively high as compared to the certainty that significant storage capacity was necessary prior to the Waste Isolation Pilot Plant (WIPP) accepting RH-TRU waste for disposal. As an alternative, the use of individual concrete overpacks provided an incremental method that could be used with the existing storage facilities and outside storage pads. For the concrete overpack concepts considered, the cylindrical design stored in a vertical orientation was determined to be the most effective

  13. Test manufacturing of copper canisters with cast inserts. Assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, C.G

    1998-08-01

    The current design of canisters for the deep repository for spent nuclear fuel consists of an outer corrosion-protective copper casing in the form of a tubular section with lid and bottom and an inner pressure-resistant insert. The insert is designed to be manufactured by casting and inside are channels in which the fuel assemblies are to be placed. Over the last years, a number of full-scale manufacturing tests of all canister components have been carried out. The purpose has been to determine and develop the best manufacturing technique and to establish long-term contacts with the best suppliers of material and technology. Part of the work has involved the developing and implementing of a quality assurance system in accordance with ISO 9001, covering the whole chain from suppliers of material up to and including the delivery of assembled canisters. This report consists of a description of the design of the canister together with current drawings and complementary technical specifications stipulating, among other things, requirements placed on different materials. The different manufacturing methods that have been used are also described and commented on in both text and illustrations. For the manufacturing of copper tubes, the roll-forming of rolled plate to tube halves and longitudinal welding is a method that has been tested on a relatively large number of tubes by now, and that probably can be developed into a functioning production method. However, the very promising outcome of performed tests on seamless tube manufacturing, has resulted in a change in direction in tube manufacturing, focusing on continued testing of extrusion as well as pierce and draw processing in the immediate future. In connection with ongoing operations, new manufacturing tests of tubes with less material thickness will be carried out. Test manufacturing of cast inserts has resulted in the choice of nodular iron as material in the continued work. This improvement in design has resulted

  14. Test manufacturing of copper canisters with cast inserts. Assessment report

    International Nuclear Information System (INIS)

    The current design of canisters for the deep repository for spent nuclear fuel consists of an outer corrosion-protective copper casing in the form of a tubular section with lid and bottom and an inner pressure-resistant insert. The insert is designed to be manufactured by casting and inside are channels in which the fuel assemblies are to be placed. Over the last years, a number of full-scale manufacturing tests of all canister components have been carried out. The purpose has been to determine and develop the best manufacturing technique and to establish long-term contacts with the best suppliers of material and technology. Part of the work has involved the developing and implementing of a quality assurance system in accordance with ISO 9001, covering the whole chain from suppliers of material up to and including the delivery of assembled canisters. This report consists of a description of the design of the canister together with current drawings and complementary technical specifications stipulating, among other things, requirements placed on different materials. The different manufacturing methods that have been used are also described and commented on in both text and illustrations. For the manufacturing of copper tubes, the roll-forming of rolled plate to tube halves and longitudinal welding is a method that has been tested on a relatively large number of tubes by now, and that probably can be developed into a functioning production method. However, the very promising outcome of performed tests on seamless tube manufacturing, has resulted in a change in direction in tube manufacturing, focusing on continued testing of extrusion as well as pierce and draw processing in the immediate future. In connection with ongoing operations, new manufacturing tests of tubes with less material thickness will be carried out. Test manufacturing of cast inserts has resulted in the choice of nodular iron as material in the continued work. This improvement in design has resulted

  15. Defense High Level Waste Disposal Container System Description

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-12

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different

  16. Neutron-sensitive ZnS/{sup 10}B{sub 2}O{sub 3} ceramic scintillator detector as an alternative to a {sup 3}He-gas-based detector for a plutonium canister assay system

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, T., E-mail: nakamura.tatsuya@jaea.go.jp [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Ohzu, A. [Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Toh, K.; Sakasai, K.; Suzuki, H.; Honda, K. [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Birumachi, A.; Ebine, M. [Nuclear Science Institute, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Yamagishi, H. [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Takase, M.; Haruyama, M.; Kureta, M. [Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Soyama, K. [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Nakamura, H. [Tokai Reprocessing Development Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Seya, M. [Integrated Support Center for Nuclear Nonproliferation and Nuclear Security, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan)

    2014-11-01

    A neutron-sensitive ZnS/{sup 10}B{sub 2}O{sub 3} ceramic scintillator detector was developed as an alternative to a {sup 3}He-gas-based detector for use in a plutonium canister assay system. The detector has a modular structure, with a flat ZnS/{sup 10}B{sub 2}O{sub 3} ceramic scintillator strip that is installed diagonally inside a light-reflecting aluminium case with a square cross-section, and where the scintillation light is detected using two photomultiplier tubes attached at both ends of the case. The prototype detectors, which have a neutron-sensitive area of 30 mm×250 mm, exhibited a sensitivity of 21.7–23.4±0.1 cps/nv (mean±SD) for thermal neutrons, a {sup 137}Cs gamma-ray sensitivity of 1.1–1.9±0.2×10{sup −7} and a count variation of less than 6% over the detector length. A trial experiment revealed a temperature coefficient of less than −0.24±0.05%/°C over the temperature range of 20–50 °C. The detector design and the experimental results are presented.

  17. Neutron-sensitive ZnS/10B2O3 ceramic scintillator detector as an alternative to a 3He-gas-based detector for a plutonium canister assay system

    Science.gov (United States)

    Nakamura, T.; Ohzu, A.; Toh, K.; Sakasai, K.; Suzuki, H.; Honda, K.; Birumachi, A.; Ebine, M.; Yamagishi, H.; Takase, M.; Haruyama, M.; Kureta, M.; Soyama, K.; Nakamura, H.; Seya, M.

    2014-11-01

    A neutron-sensitive ZnS/10B2O3 ceramic scintillator detector was developed as an alternative to a 3He-gas-based detector for use in a plutonium canister assay system. The detector has a modular structure, with a flat ZnS/10B2O3 ceramic scintillator strip that is installed diagonally inside a light-reflecting aluminium case with a square cross-section, and where the scintillation light is detected using two photomultiplier tubes attached at both ends of the case. The prototype detectors, which have a neutron-sensitive area of 30 mm×250 mm, exhibited a sensitivity of 21.7-23.4±0.1 cps/nv (mean±SD) for thermal neutrons, a 137Cs gamma-ray sensitivity of 1.1-1.9±0.2×10-7 and a count variation of less than 6% over the detector length. A trial experiment revealed a temperature coefficient of less than -0.24±0.05%/°C over the temperature range of 20-50 °C. The detector design and the experimental results are presented.

  18. Thermo-hydro-mechanical mode of canister retrieval test

    International Nuclear Information System (INIS)

    Document available in extended abstract form only. The Canister Retrieval Tests (CRT) is a full scale in situ experiment performed by SKB at Aespoe Laboratory. The experiment involves placing a canister equipped with electrical heaters inside of a deposition hole bored in Aespoe diorite. The deposition hole is 8.55 metres deep and has a diameter of 1.76 metres. The space between canister and the hole is filled with a MX-80 bentonite buffer. The bentonite buffer was installed in form of blocks and rings of bentonite. At the top of the canister bentonite bricks occupy the volume between the canister top surface and the bottom surface of the plug. Due to the bentonite ring size there are two gaps; once between canister and buffer which was left empty and another one between buffer and rock that was filled with bentonite pellets. The top of the hole was sealed with a retaining plug composed of concrete and a steel plate. The plug was secured against heave caused by the swelling clay with nine cables anchored in the rock. An artificial pressurised saturation system was used because the supply of water from the rock was judged to be insufficient for saturating the buffer in a feasible time. A large number of instruments were installed to monitor the test as follows: - Canister - temperature and strain. - Rock mass - temperature and stress. - Retaining system - force and displacement. - Buffer - temperature, relative humidity, pore pressure and total pressure. After dismantling the tests the final dry density and water content of bentonite and pellets were measured. The comprehensive record of the Thermo-Hydro-Mechanical (THM) processes in the buffer give the possibility to investigate theoretical formulations and models, since the results of THM analyses can be checked against experimental data. As part of the European project THERESA, a 2-D axisymmetric model simulation of CRT bas been carried out. Some of the main objectives of this simulation are the study of the

  19. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    A near-field performance evaluation of an Advanced Cold Process Canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. The canister design was originally proposed by TVO. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. Throughout the analysis, present day underground conditions has been assumed to persist during the service life of the canister. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie localized corrosion for the steel or copper canisters can be dismissed as a failure mechanism. The evaluation of the effects of processes outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. This factor will ensure the safety of the concept. (orig.)

  20. Development of measurement technology of chlorine attached on canister using laser. Application of LIBS using collinear geometry

    International Nuclear Information System (INIS)

    A concrete cask is adopted for interim storage of spent fuel. The facility has a natural ventilating system to cool down a stainless steel canister inside the concrete cask. When sea salt particles enter into the ventilating system and attach to the canister, the canister has a possibility to suffer SCC(Stress Corrosion Cracking) induced by chlorine. Therefore, measurement of concentration of chlorine on the canister is requested to check the occurrence of SCC. Laser-induced breakdown spectroscopy (LIBS) is suitable for on-site measurement of concentration of chlorine attached on the canister because noncontact measurement for a canister with high temperature is possible. Experiments were performed using stainless steel plates (SUS304L, SUS316L) sprayed with synthetic seawater. Nd:YAG laser beam was focused onto the SUS304L and SUS316L sample and the emission of the ablated plasma was detected by a spectrometer and an intensified CCD camera. The chlorine spectra were measured for the samples with chlorine concentration from 0.0 g/m2 to 4.0 g/m2 by using single or double pulse measurement. The double pulse measurement was designed by collinear geometry. The intensity of the chlorine fluorescence normalized by oxygen fluorescence increased monotonously versus chlorine concentration from 0.0 to 0.4 g/m2 in double pulse measurements. These results show the possibility of the quantitative measurement of chlorine content on the canister by LIBS. (author)

  1. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    Compatibility studies were conducted between 7 waste forms and 15 potential canister structural materials. The waste forms were Al-Si and Pb-Sn matrix alloys, FUETAP, glass, Synroc D, and waste particles coated with carbon or carbon plus silicon carbide. The canister materials included carbon steel (bare and with chromium or nickel coatings), copper, Monel, Cu-35% Ni, titanium (grades 2 and 12), several Inconels, aluminum alloy 5052, and two stainless steels. Tests of either 6888 or 8821 h were conducted at 100 and 3000C, which bracket the low and high limits expected during storage. Glass and FUETAP evolved sulfur, which reacted preferentially with copper, nickel, and alloys of these metals. The Pb-Sn matrix alloy stuck to all samples and the carbon-coated particles to most samples at 3000C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 8000C was included because it is representative of a transportation accident and is required of casks containing nuclear materials. During these tests (1) glass and FUETAP evolved sulfur, (2) FUETAP evolved large amounts of gas, (3) Synroc stuck to titanium alloys, (4) glass was molten, and (5) both matrix alloys were molten with considerable chemical interactions with many of the canister samples. If this test condition were imposed on waste canisters, it would be design limiting in many waste storage concepts

  2. Rehearsal: Sample Canister in Cleanroom (Animation)

    Science.gov (United States)

    2005-01-01

    [figure removed for brevity, see original site] Click on the image for Rehearsal: Sample Canister in Cleanroom animation This movie shows rehearsal of the initial processing of the sample return capsule when it is taken to a temporary cleanroom at Utah's Test and Training Range.

  3. A probabilistic approach to design of the spent fuel cask handling system of a nuclear plant

    International Nuclear Information System (INIS)

    The safety in handling containers of spent fuel assemblies, using a travelling bridge crane, during transfer to the deactivation pool and other handling operations within the fuel building, is under consideration. Probability studies of the causes of container falls which are ascribable to cranes were carried out. An analysis of the operation of the various components was followed by a determination of the frequency of such components to failure or breakdown. Available data and data collected from those industries using equipment with the same functions ie overhead travelling bridges carrying heavy loads, were qualified by analysis based on failure trees. Results and conclusions were applied to the design of a new crane and the same type of analysis was undergone. The probability of the containers then dropping proved to be below the safety standard required. (U.K.)

  4. Performance Analysis of VFD Fed Aerial Ropeway System in Coal Handling Plant at CSTPS, Chandrapur-A Case Study

    Directory of Open Access Journals (Sweden)

    Mr. Umesh G.Bonde

    2016-07-01

    Full Text Available Large rating drives required in coal handling plant for coal transportation. Smooth control with minimum power loss techniques must prefer for reliable operation. This paper presents the comparative studies of slip ring induction motor and squirrel cage induction motor with VFD technique for aerial rope way system in the view of good manageability, efficiency, and energy economy used at super thermal power station, Chandrapur, Maharashtra. Simulations are carried out in MATLAB environment

  5. DEVELOPMENT OF A TAMPER RESISTANT/INDICATING AEROSOL COLLECTION SYSTEM FOR ENVIRONMENTAL SAMPLING AT BULK HANDLING FACILITIES

    Energy Technology Data Exchange (ETDEWEB)

    Sexton, L.

    2012-06-06

    Environmental sampling has become a key component of International Atomic Energy Agency (IAEA) safeguards approaches since its approval for use in 1996. Environmental sampling supports the IAEA's mission of drawing conclusions concerning the absence of undeclared nuclear material or nuclear activities in a Nation State. Swipe sampling is the most commonly used method for the collection of environmental samples from bulk handling facilities. However, augmenting swipe samples with an air monitoring system, which could continuously draw samples from the environment of bulk handling facilities, could improve the possibility of the detection of undeclared activities. Continuous sampling offers the opportunity to collect airborne materials before they settle onto surfaces which can be decontaminated, taken into existing duct work, filtered by plant ventilation, or escape via alternate pathways (i.e. drains, doors). Researchers at the Savannah River National Laboratory and Oak Ridge National Laboratory have been working to further develop an aerosol collection technology that could be installed at IAEA safeguarded bulk handling facilities. The addition of this technology may reduce the number of IAEA inspector visits required to effectively collect samples. The principal sample collection device is a patented Aerosol Contaminant Extractor (ACE) which utilizes electrostatic precipitation principles to deposit particulates onto selected substrates. Recent work has focused on comparing traditional swipe sampling to samples collected via an ACE system, and incorporating tamper resistant and tamper indicating (TRI) technologies into the ACE system. Development of a TRI-ACE system would allow collection of samples at uranium/plutonium bulk handling facilities in a manner that ensures sample integrity and could be an important addition to the international nuclear safeguards inspector's toolkit. This work was supported by the Next Generation Safeguards Initiative (NGSI

  6. Evaluation of an Automated Liquid-Handling System (Tecan Genesis RSP 100) in the Abbott LCx Assay for Chlamydia trachomatis

    OpenAIRE

    Hanson, Kevan L.; Cartwright, Charles P.

    2001-01-01

    The present study investigated the feasibility of automating the specimen-pipetting component of sample preparation in the LCx Chlamydia assay (LCx-CT assay; Abbott Laboratories, Chicago, Ill.) by using a commercially available liquid-handling system (Tecan Genesis RSP100; Tecan Inc., Research Triangle Park, N.C.). The Tecan instrument proved to be comparable in both precision and accuracy to a manual multipipettor (Eppendorf model 4850; Eppendorf Scientific, Westbury, N.Y.). The Tecan instru...

  7. Low cost, robust and real time system for detecting and tracking moving objects to automate cargo handling in port terminals

    OpenAIRE

    Vaquero, Victor; Repiso, Ely; Sanfeliu, Alberto; Vissers, John; Kwakkernaat, Maurice

    2015-01-01

    The presented paper addresses the problem of detecting and tracking moving objects for autonomous cargo handling in port terminals using a perception system which input data is a single layer laser scanner. A computationally low cost and robust Detection and Tracking Moving Objects (DATMO) algorithm is presented to be used in autonomous guided vehicles and autonomous trucks for efficient transportation of cargo in ports. The method first detects moving objects and then tracks them, taking int...

  8. Handling time promotes the coevolution of aggregation in predator–prey systems

    OpenAIRE

    Schreiber, Sebastian J; Vejdani, Melanie

    2005-01-01

    Predators often have type II functional responses and live in environments where their life history traits as well as those of their prey vary from patch to patch. To understand how spatial heterogeneity and predator handling times influence the coevolution of patch preferences and ecological stability, we perform an ecological and evolutionary analysis of a Nicholson–Bailey type model. We prove that coevolutionarily stable prey and searching predators prefer patches that in isolation support...

  9. Fabrication plant with several working places and with a handling system

    International Nuclear Information System (INIS)

    The aim of the invention is a fabrication plant. The installation comprises several working places arranged along at least one handling line. It includes a handling device having, on one hand, an aerial carrier mobile along the handling line provided with means allowing to catch, to carry and put down a series of objects of which number is determined, and on the other hand, at right angle of each work place, an intermediate-storage and recovery device having a place A where the carrier can put down before, a place B where the carrier can recover after, and a conveyor able to take together the series of objects at the place A, to present them one by one at the working place and to put them down together at the place B, so that to decouple the carrier functioning of each working place. The invention has an important application in a nuclear fuel fabrication plant, the fuel rods comprising a tight and long can containing a stack of fuel pellets and closed by plugs

  10. A new compact and low cost Langmuir Probe and associated onboard data handling system for CubeSat

    Science.gov (United States)

    Muralikrishna, Polinaya; Domingos, Sinval; Paredes, Andres; Abrahão Dos Santos, Walter

    2016-07-01

    A new compact and low cost Langmuir Probe and associated onboard data handling system are being developed at Instituto Nacional de Pesquisas Espaciais for launching on board one of the future 2U CubeSat missions. The system is a simplified and compacted version of the Langmuir Probe payloads launched on board several Brazilian SONDA III rockets and also developed for the Brazilian scientific satellites SACI-1 and SACI-2. The onboard data handling system will have the dual functions of preprocessing the data collected by the Langmuir Probe and acting as the interface between the experiment and the on board computer. The Langmuir Probe sensor in the form of two rectangular stainless steel strips of total surface area of approximately 80cm2 will be deployed soon after the injection of the CubeSat into orbit. A sweep voltage varying linearly from 0V to 3.0V in about 1.5 seconds and then remaining fixed at 3.0V for 1 second will be applied to the LP sensor to obtain both the electron density and electron temperature. A high sensitivity preamplifier will be used to convert the sensor current expected to be in the range of a few nano amperes to a few micro amperes into a varying potential. In order to cover the large dynamic range of the expected sensor current the preamplifier output will be further amplified by a logarithmic amplifier before being sampled and sent to the data handling system. The data handling system is projected to handle 8 analog channels and 4 digital words of 8 bits each. The incoming data will be stored in a RAM and later sent to the on board computer using a serial RS422 communication protocol. The interface unit will process the telecommands received from the on board computer. The interface is also projected to do FFT analysis of the LP sensor data and send the averaged FFT spectral amplitudes in place of the original unprocessed data. The system details are presented here.

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

    International Nuclear Information System (INIS)

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

  12. Materials handling in hospitals: An analysis of how the storage and materials handling system can be improved, using the example of Hospital Santa Maria

    OpenAIRE

    Reichart, Christiane

    2010-01-01

    A Work Project, presented as part of the requirements for the Award of a Masters Degree in Management from the NOVA – School of Business and Economics Warehousing costs and especially order picking costs account for a considerable amount of total logistics costs, what leads to the purpose of this paper which is to elaborate means by which the materials handling process of hospitals can be improved. A detailed analysis of the material storage and picking process at Hospital Santa M...

  13. 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.

  14. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    The report provides a summary of the design of the canister for final disposal of nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 11 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (26 refs.)

  15. Certification of VOC canister samplers for use at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    The Waste Isolation Pilot Plant (WIPP) site is designed to demonstrate safe disposal of transuranic (TRU) mixed waste. An air monitoring program has been established at the WIPP site to verify that volatile organic compounds (VOCs) do not migrate out of the disposal unit. In this air monitoring program, modified commercially available dual canister samplers are used to collect air samples for VOC analysis. Sampler certification, sample collection, and sample analysis are performed based on the procedures contained in US Environmental Protection Agency's Compendium Method TO-14. The canister samplers are certified for cleanliness by passing humid zero air through the entire sampling system and collecting a sample in a canister over a 24-hour period. In addition, each canister sampler is certified for target compound recovery efficiency by passing a humid calibration gas standard through the entire sampling system and collecting a sample in a canister over a 24-hour period. In this paper, the authors discuss the techniques developed for meeting the stringent certification requirements of the monitoring program and present data to support the need for these stringent requirements

  16. Management of Damaged SNF Handling Operations At Paks NPP

    International Nuclear Information System (INIS)

    The issue of handling leaky fuel is one of the crucial issues of nuclear energy. It is directly connected with transportation of leaky spent nuclear fuel (SNF) that is mostly stored in the station's cooling pools. At present damaged spent nuclear fuel (SNF) of thirty VVER-440 spent fuel assemblies (SFA) is loaded into the ventilated canisters of types 28 and 29 and temporary stored in the cooling pool of Paks NPP. The report presents the milestones of preparation and safety justification of the technology for preparing canisters with damaged SNF of Paks NPP for transport to FSUE “PA “Mayak” (Russia) for reprocessing. (author)

  17. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    A near-field performance evaluation of an advanced cold process canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie local-iced corrosion for the steel or copper canisters can be dismissed as a failure mechanism; The evaluation of the effects of processed outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. For completeness also evaluation of post-failure behaviour was carried out. Analyses were focussed on low probability phenomena from faults in canisters. Some items were identified where further research is justified in order to increase knowledge of the phenomena and thus strengthen the confidence of safety margins. However, it can be concluded that the risks of these scenarios can be judged to be acceptable. This is due to the fact that firstly, the probability of occurrence of most of these scenarios can be controlled to a large extent through technical measures. Secondly, these analyses indicated that the consequences would not be severe

  18. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  19. Design report of the disposal canister for twelve fuel assemblies

    International Nuclear Information System (INIS)

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.)

  20. Multi-canister overpack closure operations location study

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel Path Forward Project (SNF Project) has been established to develop engineered methods for the expedited removal of the irradiated uranium fuel from the K East (KE) and K West (KW) Basins. As specified by the SNF Project, the SNF will be removed from the K Basins, conditioned for dry storage and placed in a long term interim storage facility located in the 200 East Area. The SNF primarily consists of Zircaloy-2 clad uranium fuel discharged from the N-Reactor. A small portion of the SNF is Single Pass Reactor (SPR) Fuel, which is aluminum clad uranium fuel. The SNF will be loaded into Multi-Canister Overpacks (MCOs) at the K Basins, transferred to the Cold Vacuum Drying (CVD) facility for initial fuel conditioning, and transported to the Canister Storage Building (CSB) for staging, final fuel conditioning, and dry storage. The MCO is a transportation, conditioning, and storage vessel. The MCO consists of a 24 inch pipe with a welded bottom closure and a top closure that is field welded after the MCO is loaded with SNF. The MCO is handled and transported in the vertical orientation during all operations. Except for operations within the CSB, the MCO is always within the transportation cask which primarily provides radiological shielding and structural protection of the MCO. The MCO closure operations location study provides a relative evaluation of location options at the K Basins and the CVD Facility and recommends that the MCO closure weld be performed, inspected, and repaired at the CVD Facility

  1. An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

    International Nuclear Information System (INIS)

    both of them) of applying the vibrational forces? 2) What is best mode of operation: first fill the canister with calcined waste and then vibrate it and refill it again, or apply vibrational forces during the filling process. By optimum or best we mean less creation of stress/strain forces during the volume reduction vibration process. Lessons learnt: This preliminary study shows that; 1) The maximum stress concentration always occurs in the canister wall, however its location varies and depends on the loading condition, and vibration process. 2) The proposed vibrational process would not cause any damages to the granulated calcined waste. 3) The first natural frequency of the longitudinal vibration of the canister is around 400 Hz, which is far away from the applied vibrational frequencies and from possibility of resonance phenomena that may cause damage to the canister 4) The relationship between the maximum internal stress and the frequency of the applied load is not parabolic. 5) The mechanical properties of the granulated calcined nuclear waste have small impact on the internal stress of the canister. Finally, the calculated data suggested that applying vibrational forces will keep the entire canister whole without any indication of development defects, and will have significant economical benefits of handling HLW and LLW in calcined forms, from waste manipulation, storage and transportation

  2. Integrated Self-Monitoring of Blood Glucose System: Handling Step Analysis

    OpenAIRE

    Freckmann, Guido; Schmid, Christina; Ruhland, Katharina; Baumstark, Annette; Haug, Cornelia

    2012-01-01

    Self-monitoring of blood glucose (SMBG) implicates a number of handling steps with the meter and the lancing device. Numerous user errors can occur during SMBG, and each step adds to the complexity of use. This report compares the required steps to perform SMBG of one fully integrated (the second generation of the Accu-Chek® Mobile), three partly integrated (Accu-Chek Compact Plus, Ascensia® Breeze®2, and Accu-Chek Aviva), and six conventional (Bayer Contour®, Bayer Contour USB, BGStar™, Free...

  3. Some propulsion system noise data handling conventions and computer programs used at the Lewis Research Center

    Science.gov (United States)

    Montegani, F. J.

    1974-01-01

    Methods of handling one-third-octave band noise data originating from the outdoor full-scale fan noise facility and the engine acoustic facility at the Lewis Research Center are presented. Procedures for standardizing, retrieving, extrapolating, and reporting these data are explained. Computer programs are given which are used to accomplish these and other noise data analysis tasks. This information is useful as background for interpretation of data from these facilities appearing in NASA reports and can aid data exchange by promoting standardization.

  4. Agro-terrorism and the Grain Handling Systems in Canada and the United States

    OpenAIRE

    Nganje, William E.; Wilson, William W.; Nolan, James F.

    2004-01-01

    The grain handling sector in Canada and the United States is vital to agriculture and trade. In a typical year on the Canadian prairies, about 140,000 producers deliver some 20 to 30 mmt of grain for export to primary elevators. In the United States, about 2.1 million producers deliver about 300 mmt of grain to primary elevators. Canadian grain is moved to export position using more than 400,000 hopper cars and marine containers, where about 1,200 ships per year are loaded. In the United Stat...

  5. RFID System Integration Design with Existing Websites via EPCglobal-like Architecture for Expensive Material Handling

    OpenAIRE

    Tenqchen, Shing; Chiu, Chui- Yu; Laraqui, Saad

    2009-01-01

    In this chapter, we propose an example of handling expensive materials using RFID technological approach on an open platform environment and follow the standardization of EPCglobal Gen II. We discuss the integration case for the case of centralized deployment. We also discuss the general cases in the future. We hope to have a good reference site for your design. The high unit price and big volume materials have an urgent request to have a clear request on input/output information needed by op...

  6. FUEL HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-06-30

    The purpose of this design calculation is to perform a criticality evaluation of the Fuel Handling Facility (FHF) and the operations and processes performed therein. The current intent of the FHF is to receive transportation casks whose contents will be unloaded and transferred to waste packages (WP) or MGR Specific Casks (MSC) in the fuel transfer bays. Further, the WPs will also be prepared in the FHF for transfer to the sub-surface facility (for disposal). The MSCs will be transferred to the Aging Facility for storage. The criticality evaluation of the FHF features the following: (I) Consider the types of waste to be received in the FHF as specified below: (1) Uncanistered commercial spent nuclear fuel (CSNF); (2) Canistered CSNF (with the exception of horizontal dual-purpose canister (DPC) and/or multi-purpose canisters (MPCs)); (3) Navy canistered SNF (long and short); (4) Department of Energy (DOE) canistered high-level waste (HLW); and (5) DOE canistered SNF (with the exception of MCOs). (II) Evaluate the criticality analyses previously performed for the existing Nuclear Regulatory Commission (NRC)-certified transportation casks (under 10 CFR 71) to be received in the FHF to ensure that these analyses address all FHF conditions including normal operations, and Category 1 and 2 event sequences. (III) Evaluate FHF criticality conditions resulting from various Category 1 and 2 event sequences. Note that there are currently no Category 1 and 2 event sequences identified for FHF. Consequently, potential hazards from a criticality point of view will be considered as identified in the ''Internal Hazards Analysis for License Application'' document (BSC 2004c, Section 6.6.4). (IV) Assess effects of potential moderator intrusion into the fuel transfer bay for defense in depth. The SNF/HLW waste transfer activity (i.e., assembly and canister transfer) that is being carried out in the FHF has been classified as safety category in the &apos

  7. Three-Dimensional Heat Transfer Analysis for A Thermal Energy Storage Canister

    Institute of Scientific and Technical Information of China (English)

    Hou Xinbin; Xin Yuming; Yang Chunxin; Yuan Xiugan; Dong Keyong

    2001-01-01

    High temperature latent thermal storage is one of the critical techniques for a solar dynamic power system. This paper presents results from heat transfer analysis of a phase change salt containment canister. A three dimensional analysis program is developed to model heat transfer in a PCM canister. Analysis include effects of asymmetric circumference heat flux, conduction in canister walls, liquid PCM and solid PCM, void volume change and void location, and conduction and radiation across PCM vapor void. The PCM phase change process is modeled using the enthalpy method and the simulation results are compared with those of other two dimensional investigations. It's shown that there are large difference with two-dimensional analysis, therefore the three-dimensional model is necessary for system design of high temperature latent thermal storage.

  8. Testing of candidate waste-package backfill and canister materials for basalt

    International Nuclear Information System (INIS)

    The Basalt Waste Isolation Project (BWIP) is developing a multiple-barrier waste package to contain high-level nuclear waste as part of an overall system (e.g., waste package, repository sealing system, and host rock) designed to isolate the waste in a repository located in basalt beneath the Hanford Site, Richland, Washington. The three basic components of the waste package are the waste form, the canister, and the backfill. An extensive testing program is under way to determine the chemical, physical, and mechanical properties of potential canister and backfill materials. The data derived from this testing program will be used to recommend those materials that most adequately perform the functions assigned to the canister and backfill

  9. Dry storage technologies: keys to choosing among metal casks, concrete shielded steel canister modules and vaults

    International Nuclear Information System (INIS)

    time. Then the key criterion is maximum modularity. Furthermore, the up front capital costs requirement for this type of solution is minimal, so depending on the chosen discount rate of the investor, they have an additional attraction. Those smaller modules allow to change course in back end policy more easily. Priority of modularity yields two other solutions, dual-purpose metal casks of the TN24TM family or dual purpose or single purpose concrete shielded welded canisters such as NUHOMS. These solutions, implemented by COGEMA LOGISTICS, TRANSNUCLEAR Inc. and FRAMATOME-ANP, are very flexible and have been adapted also to quite different fuels. Among what influences the choice, we can consider: in favor of metal casks (minimal ancillary equipment, ready to move to final or centralized repository or reprocessing or other ISFSI, compact systems, easy rearrangement, easy handling), in favor of concrete shielded canisters based systems (economics when initial quantity is sufficient to spread out up front equipment, significant cost-shielding advantage, easy local production of the relatively light canisters). Both approaches, when transportable, are also a factor for public acceptance because of the non-permanent characteristics and because transport licensing refers to internationally recognized rules, standards and methods. (authors)

  10. Effects of stabilizers on the heat transfer characteristics of a nuclear waste canister

    International Nuclear Information System (INIS)

    This report summarizes the feasibility and the effectiveness of using stabilizers (internal metal structural components) to augment the heat transfer characteristics of a nuclear waste canister. The problem was modeled as a transient two-dimensional heat transfer in two physical domains - the stabilizer and the wedge (a 30-degree-angle canister segment), which includes the heat-producing spent-fuel rods. This problem is solved by a simultaneous and interrelated numerical investigation of the two domains in cartesian and polar coordinate systems. The numerical investigations were performed for three cases. In the first case, conduction was assumed to be the dominant mechanism for heat transfer. The second case assumed that radiation was the dominant mechanism, and in the third case both radiation and conduction were considered as mechanisms of heat transfer. The results show that for typical conditions in a waste package design, the stabilizers are quite effective in reducing the overall temperature in a waste canister. Furthermore, the results show that increasing the stabilizer thickness over the thickness specified in the present design has a negligible effect on the temperature distribution in the canister. Finally, the presence of the stabilizers was found to shift the location of the peak temperature areas in the waste canister

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

  13. State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 2 - EBW

    International Nuclear Information System (INIS)

    This report consist the results of the development of the electron beam welding (EBW) method for sealing spent nuclear fuel (SNF) disposal canister. This report has been used as background material for selection of the sealing method for the SNF canister. Report contains the state of the art knowledge of the EBW method and research and development (R and D) results done by Posiva. Relevant R and D results of EB-welds done by SKB are also reviewed in this report. Requirements set for the welding and weld are present. These requirements are based on the long term safety and also some part of requirements are set by other processes like non-destructive testing (NDT) and manufacturing processes of components. Initial state of the weld is described in this report. Initial state has significant effect on the long term safety issues like corrosion resistance and creep ductility. Also short and long term mechanical properties as well as corrosion properties are described. Microstructure and residual stresses of the weld is represented in this report. Report consists also imperfections of the weld and statistical analysis of the evaluation of the probability of the largest defect size on the weld. Results of corrosion and creep tests of EB-welds are reviewed in this report. EBW process and machine are described. Preliminary designing of the EBW-machine has been done including component handling equipments. Preliminary welding procedure specification (pWPS) has drawn up and qualification of the personnel is described briefly. In-line process and quality control system including seam tracking system is implemented in modern EBW machine. Also NDT methods for inspection of the weld are described in this report. Concerning the results from the research and development work it can be concluded that EB welding method is suitable method for sealing SNF canister. Weld material fulfils requirements set by the long term safety. The welding system is robust and reliable and it is based

  14. State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 2 - EBW

    Energy Technology Data Exchange (ETDEWEB)

    Salonen, T.

    2014-05-15

    This report consist the results of the development of the electron beam welding (EBW) method for sealing spent nuclear fuel (SNF) disposal canister. This report has been used as background material for selection of the sealing method for the SNF canister. Report contains the state of the art knowledge of the EBW method and research and development (R and D) results done by Posiva. Relevant R and D results of EB-welds done by SKB are also reviewed in this report. Requirements set for the welding and weld are present. These requirements are based on the long term safety and also some part of requirements are set by other processes like non-destructive testing (NDT) and manufacturing processes of components. Initial state of the weld is described in this report. Initial state has significant effect on the long term safety issues like corrosion resistance and creep ductility. Also short and long term mechanical properties as well as corrosion properties are described. Microstructure and residual stresses of the weld is represented in this report. Report consists also imperfections of the weld and statistical analysis of the evaluation of the probability of the largest defect size on the weld. Results of corrosion and creep tests of EB-welds are reviewed in this report. EBW process and machine are described. Preliminary designing of the EBW-machine has been done including component handling equipments. Preliminary welding procedure specification (pWPS) has drawn up and qualification of the personnel is described briefly. In-line process and quality control system including seam tracking system is implemented in modern EBW machine. Also NDT methods for inspection of the weld are described in this report. Concerning the results from the research and development work it can be concluded that EB welding method is suitable method for sealing SNF canister. Weld material fulfils requirements set by the long term safety. The welding system is robust and reliable and it is based

  15. Using Single-Camera 3-D Imaging to Guide Material Handling Robots in a Nuclear Waste Package Closure System

    International Nuclear Information System (INIS)

    Nuclear reactors for generating energy and conducting research have been in operation for more than 50 years, and spent nuclear fuel and associated high-level waste have accumulated in temporary storage. Preparing this spent fuel and nuclear waste for safe and permanent storage in a geological repository involves developing a robotic packaging system--a system that can accommodate waste packages of various sizes and high levels of nuclear radiation. During repository operation, commercial and government-owned spent nuclear fuel and high-level waste will be loaded into casks and shipped to the repository, where these materials will be transferred from the casks into a waste package, sealed, and placed into an underground facility. The waste packages range from 12 to 20 feet in height and four and a half to seven feet in diameter. Closure operations include sealing the waste package and all its associated functions, such as welding lids onto the container, filling the inner container with an inert gas, performing nondestructive examinations on welds, and conducting stress mitigation. The Idaho National Laboratory is designing and constructing a prototype Waste Package Closure System (WPCS). Control of the automated material handling is an important part of the overall design. Waste package lids, welding equipment, and other tools must be moved in and around the closure cell during the closure process. These objects are typically moved from tool racks to a specific position on the waste package to perform a specific function. Periodically, these objects are moved from a tool rack or the waste package to the adjacent glovebox for repair or maintenance. Locating and attaching to these objects with the remote handling system, a gantry robot, in a loosely fixtured environment is necessary for the operation of the closure cell. Reliably directing the remote handling system to pick and place the closure cell equipment within the cell is the major challenge

  16. Investigation of Control System and Display Variations on Spacecraft Handling Qualities for Docking with Stationary and Rotating Targets

    Science.gov (United States)

    Jackson, E. Bruce; Goodrich, Kenneth H.; Bailey, Randall E.; Barnes, James R.; Ragsdale, William A.; Neuhaus, Jason R.

    2010-01-01

    This paper documents the investigation into the manual docking of a preliminary version of the Crew Exploration Vehicle with stationary and rotating targets in Low Earth Orbit. The investigation was conducted at NASA Langley Research Center in the summer of 2008 in a repurposed fixed-base transport aircraft cockpit and involved nine evaluation astronauts and research pilots. The investigation quantified the benefits of a feed-forward reaction control system thruster mixing scheme to reduce translation-into-rotation coupling, despite unmodeled variations in individual thruster force levels and off-axis center of mass locations up to 12 inches. A reduced rate dead-band in the phase-plane attitude controller also showed some promise. Candidate predictive symbology overlaid on a docking ring centerline camera image did not improve handling qualities, but an innovative attitude status indicator symbol was beneficial. The investigation also showed high workload and handling quality problems when manual dockings were performed with a rotating target. These concerns indicate achieving satisfactory handling quality ratings with a vehicle configuration similar to the nominal Crew Exploration Vehicle may require additional automation.

  17. Measurements of Fundamental Fluid Physics of SNF Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Condie, Keith Glenn; Mc Creery, Glenn Ernest; McEligot, Donald Marinus

    2001-09-01

    With the University of Idaho, Ohio State University and Clarksean Associates, this research program has the long-term goal to develop reliable predictive techniques for the energy, mass and momentum transfer plus chemical reactions in drying / passivation (surface oxidation) operations in the transfer and storage of spent nuclear fuel (SNF) from wet to dry storage. Such techniques are needed to assist in design of future transfer and storage systems, prediction of the performance of existing and proposed systems and safety (re)evaluation of systems as necessary at later dates. Many fuel element geometries and configurations are accommodated in the storage of spent nuclear fuel. Consequently, there is no one generic fuel element / assembly, storage basket or canister and, therefore, no single generic fuel storage configuration. One can, however, identify generic flow phenomena or processes which may be present during drying or passivation in SNF canisters. The objective of the INEEL tasks was to obtain fundamental measurements of these flow processes in appropriate parameter ranges.

  18. Pressurization of whole element canister during staging

    International Nuclear Information System (INIS)

    An analytical model was developed to estimate the buildup of gas pressure for a single outer element in a hot cell test container for a post cold vacuum drying staging/storage test. This model considers various sources of gas generation and gas consumption as a function of time. In a canister containing spent nuclear fuel, hydrogen is generated from the reactions of uranium with free water or hydrated water, hydride decomposition, and radiolysis. The canister pressurization model predicts a stable pressure and a peak temperature during staging, with an assumption that a fuel element contains 40 gm of corrosion products and a decay heat of 2.07 or 1.06 Watts. Calculations were also performed on constant temperature tests for fuel elements containing varied amounts of sludge tested at 150, 125, 105, and 85 C. The pressurization model will be used to evaluate test results obtained from post-drying testing on whole fuel elements

  19. Multi-canister overpack: additional NRC requirements

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) established in the K Basin Spent Fuel Project, Regulatory Policy, dated August 4, 1995 (hereafter referred to as the Policy), the requirement for new Spent Nuclear Fuel Project (SNFP) facilities to achieve ''nuclear safety equivalency'' to comparable U.S. Nuclear Regulatory Commission licensed facilities. For activities other than during transport, when the Multi-Canister Overpack (MCO) is used and resides in the Canister Storage Building (CSB), Conditioning Facility or K Basins Path Forward Projects, additional NRC requirements will also apply to the MCO based on the safety functions it performs and its interfaces with the SNFP facilities. An evaluation was performed in consideration of the MCO safety functions to identify any additional NRC requirements, to establish nuclear safety equivalency for the MCO

  20. Oxidative Dissolution of Spent Fuel and Release of Nuclides from a Copper/Iron Canister : Model Developments and Applications

    OpenAIRE

    Liu, Longcheng

    2001-01-01

    Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechani...