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Sample records for canister handling systems

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

    International Nuclear Information System (INIS)

    Wendelin, T.; Suikki, M.

    2008-12-01

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

  2. CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    B. Gorpani

    2000-01-01

    The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane hoist,; DC--loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the; DC--is fully loaded, the Disposal Container Transport System moves the; DC--to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister

  3. Canister Transfer System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

    The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane/hoist, DC loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the DC is fully loaded, the Disposal Container Transport System moves the DC to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister handling

  4. 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.

  5. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    Beesley. J.F.

    2005-01-01

    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

  6. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    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

  7. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    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

  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. Thermal studies of the canister staging pit in a hypothetical Yucca Mountain canister handling facility using computational fluid dynamics

    International Nuclear Information System (INIS)

    Soltani, Mehdi; Barringer, Chris; Bues, Timothy T. de

    2007-01-01

    The proposed Yucca Mountain nuclear waste storage site will contain facilities for preparing the radioactive waste canisters for burial. A previous facility design considered was the Canister Handling Facility Staging Pit. This design is no longer used, but its thermal evaluation is typical of such facilities. Structural concrete can be adversely affected by the heat from radioactive decay. Consequently, facilities must have heating ventilation and air conditioning (HVAC) systems for cooling. Concrete temperatures are a function of conductive, convective and radiative heat transfer. The prediction of concrete temperatures under such complex conditions can only be adequately handled by computational fluid dynamics (CFD). The objective of the CFD analysis was to predict concrete temperatures under normal and off-normal conditions. Normal operation assumed steady state conditions with constant HVAC flow and temperatures. However, off-normal operation was an unsteady scenario which assumed a total HVAC failure for a period of 30 days. This scenario was particularly complex in that the concrete temperatures would gradually rise, and air flows would be buoyancy driven. The CFD analysis concluded that concrete wall temperatures would be at or below the maximum temperature limits in both the normal and off-normal scenarios. While this analysis was specific to a facility design that is no longer used, it demonstrates that such facilities are reasonably expected to have satisfactory thermal performance. (author)

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

    International Nuclear Information System (INIS)

    1994-09-01

    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

  11. The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

  12. Preliminary Transportation, Aging and Disposal Canister System Performance Specification

    International Nuclear Information System (INIS)

    C.A Kouts

    2006-01-01

    This document provides specifications for selected system components of the Transportation, Aging and Disposal (TAD) canister-based system. A list of system specified components and ancillary components are included in Section 1.2. The TAD canister, in conjunction with specialized overpacks will accomplish a number of functions in the management and disposal of spent nuclear fuel. Some of these functions will be accomplished at purchaser sites where commercial spent nuclear fuel (CSNF) is stored, and some will be performed within the Office of Civilian Radioactive Waste Management (OCRWM) transportation and disposal system. This document contains only those requirements unique to applications within Department of Energy's (DOE's) system. DOE recognizes that TAD canisters may have to perform similar functions at purchaser sites. Requirements to meet reactor functions, such as on-site dry storage, handling, and loading for transportation, are expected to be similar to commercially available canister-based systems. This document is intended to be referenced in the license application for the Monitored Geologic Repository (MGR). As such, the requirements cited herein are needed for TAD system use in OCRWM's disposal system. This document contains specifications for the TAD canister, transportation overpack and aging overpack. The remaining components and equipment that are unique to the OCRWM system or for similar purchaser applications will be supplied by others

  13. Shippingport Spent Fuel Canister System Description

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    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

  14. Remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility is being constructed at the Savannah River Plant at a cost of $870 million to immobilize the defense high-level radioactive waste. This radioactive waste is being added to borosilicate glass for later disposal in a federal repository. The borosilicate glass is poured into stainless steel canisters for storage. These canisters must be handled remotely because of their high radioactivity, up to 5000 R/h. After the glass has been poured into the canister which will be temporarily sealed, it is transferred to a decontamination cell and decontaminated. The canister is then transferred to the weld cell where a permanent cap is welded into place. The canisters must then be transported from the processing building to a storage vault on the plant until the federal repository is available. A shielded canister transporter (SCT) has been designed and constructed for this purpose. The design of the SCT vehicle allows the safe transport of a highly radioactive canister containing borosilicate glass weighing 2300 kg with a radiation level up to 5000 R/h from one building to another. The design provides shielding for the operator in the cab of the vehicle to be below 0.5 rem/h. The SCT may also be used to load the final shipping cask when the federal repository is ready to receive the canisters

  15. A welding system for spent fuel canister lid

    International Nuclear Information System (INIS)

    Suikki, M.; Wendelin, T.

    2008-06-01

    The report presents a proposed welding system for spent fuel canister lids. The system is used for welding the copper lid to the copper overpack. The apparatus will be installed in the encapsulation plant. The report presents basic requirements for and implementation of the welding system, operation, service and maintenance of the equipment, as well as a cost estimate. Some aspects of the apparatus design are quite specified, but the actual detailed planning and final selection of components is not included. The report also describes actions for possible malfunction and fault conditions. Closing of the copper cylinder's lid is carried out by electron beam welding, which must be performed in vacuum. The welding system for spent fuel canister lid consists of two welding chambers, a canister docking system, an EB-welding machine with its accessories, a vacuum apparatus, as well as necessary auxiliary equipment. The system's equipment is housed in a welding room, an auxiliary system room, an operation control room, as well as mounted on the ceiling of a transfer corridor. One of the welding chambers is intended for carrying out test welding procedures and for calibration of welding parameters. The actual spent fuel canister lid welding chamber has a weldingready canister docked thereto in an airtight manner. The chamber is pumped for a vacuum, followed by closing the canister's copper lid and carrying out the lid welding process. The lid is brought into the chamber prior to docking the canister by means of a canister transfer trolley lifting gear. Lifting of the canister and rotating it during a welding process are also handled by means of the transfer trolley. The lid welding chamber houses equipment for the alignment and installation of the lid, as well as heating means for the top side of a copper overpack for ensuring a sufficient installation clearance between the lid and the overpack. The equipment not needed in the immediate vicinity of welding chambers, is

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Herlan Martono; Aisyah; Wati

    2007-01-01

    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/mm 2 . (author)

  19. Design of double containment canister cask storage system

    International Nuclear Information System (INIS)

    Asami, M.; Matsumoto, T.; Oohama, T.; Kuriyama, K.; Kawakami, K.

    2004-01-01

    Spent fuels discharged from Japanese LWR will be stored as recycled-fuel-resources in interim storage facilities. The concrete cask storage system is one of important forms for the spent fuel interim storage. In Japan, the interim storage facility will be located near the coast, therefore it is important to prevent SCC (Stress Corrosion Cracking) caused by sea salt particles and to assure the containment integrity of the canister which contains spent fuels. KEPCO, NFT and OCL have designed the double containment canister cask storage system that can assure the long-term containment integrity and monitor the containment performance without storage capacity decrease. Major features of the combined canister cask system are shown as follows: This system can survey containment integrity of dual canisters by monitoring the pressure of the gap between canisters. The primary canister has dual lids sealed by welding. The secondary canister has single lid tightened by bolts and sealed by metallic gaskets. The primary canister is contained in the transport cask during transportation, and the gap between the primary canister and the transport cask is filled with He gas. Under storage condition in the concrete cask, the primary canister is contained in the secondary canister, and the gap between these canisters is filled with helium gas. Hence this system can prevent the primary canister to contact sea salt particle in the air and from SCC. Decrease of cooling performance because of the double canister is compensated by fins fitted on the secondary canister surface. Then, this system can prevent the decrease of storage capacity determined by the fuel temperature limit. This system can assure that the primary canister will keep intact for long term storage. Therefore, in the case of pressure down of the gap between canisters, it can be considered that the secondary canister containment is damaged, and the primary canister will be transferred to another secondary canister at the

  20. 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

  1. Stakeholder involvement in the evaluation of a multipurpose canister system

    International Nuclear Information System (INIS)

    Williams, J.R.; Kane, D.; Smith, T.B. Jr.

    1994-01-01

    The U.S. Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM), began evaluating a multipurpose canister (MPC) concept in October of 1992. This followed recommendations by the Nuclear Waste Technical Review Board (NWTRB) and the U.S. Nuclear Regulatory Commission (NRC) that DOE develop a nuclear waste management system that achieves system integration, standardization, and reduced fuel-handling operations. Industry organizations such as Edison Electric Institute (EEI) and Electric Power Research Institute (EPRI) had conducted earlier studies that concluded advantages to the nuclear waste management system may be offered by such a concept. The MPC concept involves a metal canister which would contain multiple spent nuclear fuel assemblies. The canister would be sealed at the nuclear power plant and would not be reopened. The MPC would then be placed inside separate casks or overpacks for storage, transportation, and disposal. An important factor in DOE's evaluation of the MPC concept was the involvement of external parties. This paper describes that involvement process for the OCRWM's MPC implementation program. External parties who have an interest or stake in the program are referred to as stakeholders

  2. Structural performance of a multipurpose canister shell for HLNW under normal handling conditions

    International Nuclear Information System (INIS)

    Ladkany, S.G.; Rajagopalan, R.

    1994-01-01

    A Multipurpose Canister (MPC) is analyzed for critical stresses that occur during normal handling conditions and accidental scenarios. Linear and Non-linear Finite Element Analysis is performed and the stresses at various critical locations in the MPC and its weldments are studied extensively. Progressive failure analysis of the MPC's groove and fillet welds, is presented. The structural response of the MPC to dynamic lifting loads, to loads resulting from an accidental slippage of a crane cable carrying the MPC, and from the impact between two canisters, is evaluated. Nonlinear structural analysis is used in the evaluation of the local buckling and the ultimate failure phenomena in the shell when the steel is in the strain hardening state during impact. Results make a case for increasing the thickness of the shell and all the welds

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

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

  4. System for handling and storing radioactive waste

    Science.gov (United States)

    Anderson, John K.; Lindemann, Paul E.

    1984-01-01

    A system and method for handling and storing spent reactor fuel and other solid radioactive waste, including canisters to contain the elements of solid waste, storage racks to hold a plurality of such canisters, storage bays to store these racks in isolation by means of shielded doors in the bays. This system also includes means for remotely positioning the racks in the bays and an access tunnel within which the remotely operated means is located to position a rack in a selected bay. The modular type of these bays will facilitate the construction of additional bays and access tunnel extension.

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  6. Electropolishing decontamination system for high-level waste canisters

    International Nuclear Information System (INIS)

    Larson, D.E.; Berger, D.N.; Allen, R.P.; Bryan, G.H.; Place, B.G.

    1988-10-01

    As part of a US Department of Energy (DOE) project agreement with the Federal Ministry for Research and Technology (BMFT) in the Federal Republic of Germany (FRG). The Nuclear Waste Treatment Program at the Pacific Northwest Laboratory (PNL) is preparing 30 radioactive canisters containing borosilicate glass for use in high-level waste repository related tests at the Asse Salt Mine. After filling, the canisters will be welded closed and decontaminated in preparation for shipping to the FRG. Electropolishing was selected as the primary decontamination approach, and an electropolishing system with associated canister inspection equipment has been designed and fabricated for installation in a large hot cell. This remote electropolishing system, which is currently undergoing preliminary testing, is described in this report. 3 refs., 3 figs., 1 tab

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

    International Nuclear Information System (INIS)

    FARWICK, C.C.

    1999-01-01

    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

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    FARWICK, C.C.

    2000-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-04-13

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

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

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2000-01-01

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

  12. Torus sector handling system

    International Nuclear Information System (INIS)

    Grisham, D.L.

    1981-01-01

    A remote handling system is proposed for moving a torus sector of the accelerator from under the cryostat to a point where it can be handled by a crane and for the reverse process for a new sector. Equipment recommendations are presented, as well as possible alignment schemes. Some general comments about future remote-handling methods and the present capabilities of existing systems will also be included. The specific task to be addressed is the removal and replacement of a 425 to 450 ton torus sector. This requires a horizontal movement of approx. 10 m from a normal operating position to a point where its further transport can be accomplished by more conventional means (crane or floor transporter). The same horizontal movement is required for reinstallation, but a positional tolerance of 2 cm is required to allow reasonable fit-up for the vacuum seal from the radial frames to the torus sector. Since the sectors are not only heavy but rather tall and narrow, the transport system must provide a safe, stable, and repeatable method fo sector movement. This limited study indicates that the LAMPF-based method of transporting torus sectors offers a proven method of moving heavy items. In addition, the present state of the art in remote equipment is adequate for FED maintenance

  13. MFTF exception handling system

    International Nuclear Information System (INIS)

    Nowell, D.M.; Bridgeman, G.D.

    1979-01-01

    In the design of large experimental control systems, a major concern is ensuring that operators are quickly alerted to emergency or other exceptional conditions and that they are provided with sufficient information to respond adequately. This paper describes how the MFTF exception handling system satisfies these requirements. Conceptually exceptions are divided into one of two classes. Those which affect command status by producing an abort or suspend condition and those which fall into a softer notification category of report only or operator acknowledgement requirement. Additionally, an operator may choose to accept an exception condition as operational, or turn off monitoring for sensors determined to be malfunctioning. Control panels and displays used in operator response to exceptions are described

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

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

  15. Shielded Canister Transporter

    International Nuclear Information System (INIS)

    Eidem, G.G. Jr.; Fages, R.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) will produce canisters filled with high-level radioactive waste immobilized in borosilicate glass. This report discusses a Shielded Canister Transporter (SCT) which will provide the means for safe transportation and handling of the canisters from the Vitrification Building to the Canister Storage Building (CSB). The stainless steel canisters are 0.61 meters in diameter, 3.0 meters tall, and weigh approximately 2,135 kilograms, with a maximum exterior surface dose rate of 90,000 R/hr. The canisters are placed into storage tubes to a maximum of three tall (two for overpack canisters) with an impact limiter placed at the tube bottom and between each canister. A floor plug seals the top of the storage tube at the operating floor level of the CSB

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

    International Nuclear Information System (INIS)

    Morissette, R.P.; Schneringer, P.E.; Lane, R.K.; Moore, R.L.; Young, K.A.

    1986-01-01

    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

  17. Sophisticated fuel handling system evolved

    International Nuclear Information System (INIS)

    Ross, D.A.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

    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)

    JOHNSON, D.M.

    2000-01-01

    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. Test plan for the Sample Transfer Canister system

    International Nuclear Information System (INIS)

    Flanagan, B.D.

    1998-01-01

    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. These samples will be packaged and shipped in accordance with the US Department of Transportation (DOT) regulation 49 CFR 173.4, ''Exceptions for small quantities.'' The Sample Transfer Canister is of a ''French Can'' design, intended to be mated with a glove box for loading/unloading. Transport will typically take place north of the Wye Barricade between WRAP and the 222-S Laboratory. The Sample Transfer Canister will be shipped in an insulated ice chest, but the ice chest will not be a part of the small quantity package during prototype testing

  1. 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)

    McClelland, B.C.; Moore, T.D.

    2006-01-01

    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)

  2. Zero-Headspace Coal-Core Gas Desorption Canister, Revised Desorption Data Analysis Spreadsheets and a Dry Canister Heating System

    Science.gov (United States)

    Barker, Charles E.; Dallegge, Todd A.

    2005-01-01

    Coal desorption techniques typically use the U.S. Bureau of Mines (USBM) canister-desorption method as described by Diamond and Levine (1981), Close and Erwin (1989), Ryan and Dawson (1993), McLennan and others (1994), Mavor and Nelson (1997) and Diamond and Schatzel (1998). However, the coal desorption canister designs historically used with this method have an inherent flaw that allows a significant gas-filled headspace bubble to remain in the canister that later has to be compensated for by correcting the measured desorbed gas volume with a mathematical headspace volume correction (McLennan and others, 1994; Mavor and Nelson, 1997).

  3. DISPOSAL CONTAINER HANDLING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    E. F. Loros

    2000-06-30

    The Disposal Container Handling System receives and prepares new disposal containers (DCs) and transfers them to the Assembly Transfer System (ATS) or Canister Transfer System (CTS) for loading. The system receives the loaded DCs from ATS or CTS and welds the lids. When the welds are accepted the DCs are termed waste packages (WPs). The system may stage the WP for later transfer or transfer the WP directly to the Waste Emplacement/Retrieval System. The system can also transfer DCs/WPs to/from the Waste Package Remediation System. The Disposal Container Handling System begins with new DC preparation, which includes installing collars, tilting the DC upright, and outfitting the container for the specific fuel it is to receive. DCs and their lids are staged in the receipt area for transfer to the needed location. When called for, a DC is put on a cart and sent through an airlock into a hot cell. From this point on, all processes are done remotely. The DC transfer operation moves the DC to the ATS or CTS for loading and then receives the DC for welding. The DC welding operation receives loaded DCs directly from the waste handling lines or from interim lag storage for welding of the lids. The welding operation includes mounting the DC on a turntable, removing lid seals, and installing and welding the inner and outer lids. After the weld process and non-destructive examination are successfully completed, the WP is either staged or transferred to a tilting station. At the tilting station, the WP is tilted horizontally onto a cart and the collars removed. The cart is taken through an air lock where the WP is lifted, surveyed, decontaminated if required, and then moved into the Waste Emplacement/Retrieval System. DCs that do not meet the welding non-destructive examination criteria are transferred to the Waste Package Remediation System for weld preparation or removal of the lids. The Disposal Container Handling System is contained within the Waste Handling Building System

  4. Application of plutonium inventory measurement system (PIMS) and temporary canister verification system (TCVS) at RRP

    International Nuclear Information System (INIS)

    Noguchi, Yoshihiko; Nakamura, Hironobu; Adachi, Hideto; Iwamoto, Tomonori

    2004-01-01

    In U-Pu co-denitration area at Rokkasho Reprocessing Plant (RRP), Plutonium Inventory Measurement System (PIMS) and Temporary Canister Verification System (TCVS) are installed to provide efficient and effective safeguards. PIMS measures Pu quantity inside pipes and vessels installed in glove boxes by total neutron counting method. PIMS consists of total 142 neutron detector attached on the wall and top of glove boxes and neutron count rates of each detectors are related to each other to calculate Pu quantity of each process areas. In this moment, inactive calibration using Cf-source was completed. On the other hand, TCVS measures Pu quantity of canisters inside temporary storage by coincidence counting method and it will be installed before the active test. These systems have monitoring function as additional measures. This paper describes specification, performance and measurement principles of PIMS and TCVS. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Burdin, J.R.

    1995-10-25

    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

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

    International Nuclear Information System (INIS)

    Burdin, J.R.

    1995-01-01

    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

  7. HLW Canister and Can-In-Canister Drop Calculation

    International Nuclear Information System (INIS)

    H. Marr

    1999-01-01

    The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver

  8. Safeguards considerations related to the use of multi-purpose canisters in the Civilian Radioactive Waste Management system

    International Nuclear Information System (INIS)

    Floyd, W.C.

    1995-01-01

    The US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is responsible for disposing of the nation's high-level radioactive waste. Currently, DOE is considering the use of Multi-Purpose Canisters (MPCs) to containerize commercial spent nuclear fuel (SNF) to be handled by the system. To achieve its safeguards and security objectives, OCRWM plans to institute a US Regulatory Commission (NRC)-approved safeguards program. Since the Mined Geologic Disposal System (MGDS) facility and a possible Monitored Retrievable Storage (MRS) facility may be subject to selection for International Atomic Energy Agency (IAEA) inspections, the safeguards program for MPCs may not preclude compliance with the requirements of the IAEA's Annex D, Special Criteria for Difficult-to-Access Fuel Items. MPC safeguards are based on three principles: Verification, Material Control and Accounting, and Physical Protection

  9. ATA diagnostic data handling system: an overview

    International Nuclear Information System (INIS)

    Chambers, F.W.; Kallman, J.; McDonald, J.; Slominski, M.

    1984-01-01

    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

  10. 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

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

    International Nuclear Information System (INIS)

    Ryu, Yong Ho

    1992-02-01

    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

  12. Spent fuel canister docking station

    International Nuclear Information System (INIS)

    Suikki, M.

    2006-01-01

    The working report for the spent fuel canister docking station presents a design for the operation and structure of the docking equipment located in the fuel handling cell for the spent fuel in the encapsulation plant. The report contains a description of the basic requirements for the docking station equipment and their implementation, the operation of the equipment, maintenance and a cost estimate. In the designing of the equipment all the problems related with the operation have been solved at the level of principle, nevertheless, detailed designing and the selection of final components have not yet been carried out. In case of defects and failures, solutions have been considered for postulated problems, and furthermore, the entire equipment was gone through by the means of systematic risk analysis (PFMEA). During the docking station designing we came across with needs to influence the structure of the actual disposal canister for spent nuclear fuel, too. Proposed changes for the structure of the steel lid fastening screw were included in the report. The report also contains a description of installation with the fuel handling cell structures. The purpose of the docking station for the fuel handling cell is to position and to seal the disposal canister for spent nuclear fuel into a penetration located on the cell floor and to provide suitable means for executing the loading of the disposal canister and the changing of atmosphere. The designed docking station consists of a docking ring, a covering hatch, a protective cone and an atmosphere-changing cap as well as the vacuum technology pertaining to the changing of atmosphere and the inert gas system. As far as the solutions are concerned, we have arrived at rather simple structures and most of the actuators of the system are situated outside of the actual fuel handling cell. When necessary, the equipment can also be used for the dismantling of a faulty disposal canister, cut from its upper end by machining. The

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

    International Nuclear Information System (INIS)

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

    1997-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

  15. Cask system design guidance for robotic handling

    International Nuclear Information System (INIS)

    Griesmeyer, J.M.; Drotning, W.D.; Morimoto, A.K.; Bennett, P.C.

    1990-10-01

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

  16. Sequence trajectory generation for garment handling systems

    OpenAIRE

    Liu, Honghai; Lin, Hua

    2008-01-01

    This paper presents a novel generic approach to the planning strategy of garment handling systems. An assumption is proposed to separate the components of such systems into a component for intelligent gripper techniques and a component for handling planning strategies. Researchers can concentrate on one of the two components first, then merge the two problems together. An algorithm is addressed to generate the trajectory position and a clothes handling sequence of clothes partitions, which ar...

  17. Numerical analysis of a natural convection cooling system for radioactive canisters storage

    Energy Technology Data Exchange (ETDEWEB)

    Tsal, R.J.; Anwar, S.; Mercada, M.G. [Fluor Daniel Inc., Irvine, CA (United States)

    1995-02-01

    This paper describes the use of numerical analysis for studying natural convection cooling systems for long term storage of heat producing radioactive materials, including special nuclear materials and nuclear waste. The paper explains the major design philosophy, and shares the experiences of numerical modeling. The strategy of storing radioactive material is to immobilize nuclear high-level waste by a vitrification process, convertion it into borosilicate glass, and cast the glass into stainless steel canisters. These canisters are seal welded, decontaminated, inspected, and temporarily stored in an underground vault until they can be sent to a geologic repository for permanent storage. These canisters generate heat by nuclear decay of radioactive isotopes. The function of the storage facility ventilation system is to ensure that the glass centerline temperature does not exceed the glass transition temperature during storage and the vault concrete temperatures remain within the specified limits. A natural convection cooling system was proposed to meet these functions. The effectiveness of a natural convection cooling system is dependent on two major factors that affect air movement through the vault for cooling the canisters: (1) thermal buoyancy forces inside the vault which create a stack effect, and (2) external wind forces, that may assist or oppose airflow through the vault. Several numerical computer models were developed to analyze the thermal and hydraulic regimes in the storage vault. The Site Model is used to simulate the airflow around the building and to analyze different air inlet/outlet devices. The Airflow Model simulates the natural convection, thermal regime, and hydraulic resistance in the vault. The Vault Model, internal vault temperature stratification; and, finally, the Hot Area Model is used for modeling concrete temperatures within the vault.

  18. Transportation system benefits of early deployment of a 75-ton multipurpose canister system

    International Nuclear Information System (INIS)

    Wankerl, M.W.; Schmid, S.P.

    1995-01-01

    In 1993 the US Civilian Radioactive Waste Management System (CRWMS) began developing two multipurpose canister (MPC) systems to provide a standardized method for interim storage and transportation of spent nuclear fuel (SNF) at commercial nuclear power plants. One is a 75-ton concept with an estimated payload of about 6 metric tons (t) of SNF, and the other is a 125-ton concept with an estimated payload of nearly 11 t of SNF. These payloads are two to three times the payloads of the largest currently certified US rail transport casks, the IF-300. Although is it recognized that a fully developed 125-ton MPC system is likely to provide a greater cost benefit, and radiation exposure benefit than the lower-capacity 75-ton MPC, the authors of this paper suggest that development and deployment of the 75-ton MPC prior to developing and deploying a 125-ton MPC is a desirable strategy. Reasons that support this are discussed in this paper

  19. 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.

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

    International Nuclear Information System (INIS)

    Hoekmark, Harald

    2003-11-01

    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

  1. Data handling systems and methods of wiring

    International Nuclear Information System (INIS)

    Grant, J.

    1981-01-01

    An improved data handling system, for monitoring and control of nuclear reactor operations, is described in which time delays associated with scanning are reduced and noise and fault signals in the system are resolved. (U.K.)

  2. 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.)

  3. Remote controlled mover for disposal canister transfer

    International Nuclear Information System (INIS)

    Suikki, M.

    2013-10-01

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    2007-09-15

    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

  5. 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

  6. Automated system for handling tritiated mixed waste

    International Nuclear Information System (INIS)

    Dennison, D.K.; Merrill, R.D.; Reitz, T.C.

    1995-03-01

    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

  7. Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2016-01-01

    Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

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

    International Nuclear Information System (INIS)

    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: (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

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

    International Nuclear Information System (INIS)

    GARRISON, R.C.

    2000-01-01

    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

  10. Handling system for nuclear fuel pellet inspection

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  11. MHSS: a material handling system simulator

    Energy Technology Data Exchange (ETDEWEB)

    Pomernacki, L.; Hollstien, R.B.

    1976-04-07

    A Material Handling System Simulator (MHSS) program is described that provides specialized functional blocks for modeling and simulation of nuclear material handling systems. Models of nuclear fuel fabrication plants may be built using functional blocks that simulate material receiving, storage, transport, inventory, processing, and shipping operations as well as the control and reporting tasks of operators or on-line computers. Blocks are also provided that allow the user to observe and gather statistical information on the dynamic behavior of simulated plants over single or replicated runs. Although it is currently being developed for the nuclear materials handling application, MHSS can be adapted to other industries in which material accountability is important. In this paper, emphasis is on the simulation methodology of the MHSS program with application to the nuclear material safeguards problem. (auth)

  12. PREPD O and VE remote handling system

    International Nuclear Information System (INIS)

    Theil, T.N.

    1985-01-01

    The Process Experimental Pilot Plant (PREPP) at the Idaho National Engineering Laboratory is designed for volume reduction and packaging of transuranic (TRU) waste. The PREPP opening and verification enclosure (O and VE) remote handling system, within that facility, is designed to provide examination of the contents of various TRU waste storage containers. This remote handling system will provide the means of performing a hazardous operation that is currently performed manually. The TeleRobot to be used in this system is a concept that will incorporate and develop man in the loop operation (manual mode), standardized automatic sequencing of end effector tools, increased payload and reach over currently available computer-controlled robots, and remote handling of a hazardous waste operation. The system is designed within limited space constraints and an operation that was originally planned, and is currently being manually performed at other plants. The PREPP O and VE remote handling system design incorporates advancing technology to improve the working environment in the nuclear field

  13. Multi-purpose canisters as an alternative for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Hollaway, W.R.; Rozier, R.; Nitti, D.A.; Williams, J.R.

    1993-01-01

    A study was conducted to assess the feasibility of using multi-purpose canisters to handle spent nuclear fuel throughout the Civilian Radioactive Waste Management System. Multi-purpose canisters would be sealed, metallic containers maintaining multiple spent fuel assemblies in a dry, inert environment and overpacked separately and uniquely for the various system elements of storage, transportation, and disposal. Using five implementation scenarios, the multi-purpose canister was evaluated with regard to several measures of effectiveness, including number of handlings, radiation exposure, cost, schedule and licensing considerations, and public perception. Advantages and disadvantages of the multi-purpose canister were identified relative to the current reference system within each scenario, and the scenarios were compared to determine the most effective method of implementation

  14. 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.

  15. 2D and 3D thermal simulations for storage systems with internal natural convection for canistered spent fuel

    International Nuclear Information System (INIS)

    Yaksh, M.; Wang, C.

    2004-01-01

    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

  16. Inert gas handling in ion plating systems

    International Nuclear Information System (INIS)

    Goode, A.R.; Burden, M.St.J.

    1979-01-01

    The results of an investigation into the best methods for production and monitoring of the inert gas environment for ion plating systems are reported. Work carried out on Pirani gauges and high pressure ion gauges for the measurement of pressures in the ion plating region (1 - 50mtorr) and the use of furnaces for cleaning argon is outlined. A schematic of a gas handling system is shown and discussed. (UK)

  17. VVER NPPs fuel handling machine control system

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. Status of the Multipurpose Canister (MPC) Project

    International Nuclear Information System (INIS)

    Hopper, J.P.

    1996-01-01

    The multipurpose canister (MPC) project represents a cornerstone of the current U.S. Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) program for handling spent nuclear fuel. The MPC and associated support equipment is being designed to accommodate the requirements for not only storage and transport but also for the specified disposal requirements of the mined geologic repository system. The phase 1 design effort for the MPC system, being performed by the Westinghouse team on behalf of TRW Environmental Safety Systems (TESS), the OCRWM management ampersand operating (M ampersand O) contractor, is on schedule for delivery of completed safety analysis reports (SARs) in April 1996

  19. Remote handling systems for the Pride application

    International Nuclear Information System (INIS)

    Kim, K.; Lee, J.; Lee, H.; Kim, S.; Kim, H.

    2010-10-01

    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)

  20. CLASSIFICATION OF THE MGR MUCK HANDLING SYSTEM

    International Nuclear Information System (INIS)

    R. Garrett

    1999-01-01

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) muck 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). This QA classification incorporates the current MGR design and the results of the ''Preliminary Preclosure Design Basis Event Calculations for the Monitored Geologic Repository (CRWMS M and O 1998a)

  1. DISPOSABLE CANISTER WASTE ACCEPTANCE CRITERIA

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2001-07-30

    The purpose of this calculation is to provide the bases for defining the preclosure limits on radioactive material releases from radioactive waste forms to be received in disposable canisters at the Monitored Geologic Repository (MGR) at Yucca Mountain. Specifically, this calculation will provide the basis for criteria to be included in a forthcoming revision of the Waste Acceptance System Requirements Document (WASRD) that limits releases in terms of non-isotope-specific canister release dose-equivalent source terms. These criteria will be developed for the Department of Energy spent nuclear fuel (DSNF) standard canister, the Multicanister Overpack (MCO), the naval spent fuel canister, the High-Level Waste (HLW) canister, the plutonium can-in-canister, and the large Multipurpose Canister (MPC). The shippers of such canisters will be required to demonstrate that they meet these criteria before the canisters are accepted at the MGR. The Quality Assurance program is applicable to this calculation. The work reported in this document is part of the analysis of DSNF and is performed using procedure AP-3.124, Calculations. The work done for this analysis was evaluated according to procedure QAP-2-0, Control of Activities, which has been superseded by AP-2.21Q, Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities. This evaluation determined that such activities are subject to the requirements of DOE/RW/0333P, Quality Assurance Requirements and Description (DOE 2000). This work is also prepared in accordance with the development plan titled Design Basis Event Analyses on DOE SNF and Plutonium Can-In-Canister Waste Forms (CRWMS M&O 1999a) and Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages (CRWMS M&O 2000d). This calculation contains no electronic data applicable to any electronic data management system.

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

    International Nuclear Information System (INIS)

    Suikki, M.

    2013-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    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. Remote-handled transuranic system assessment appendices. Volume 2

    International Nuclear Information System (INIS)

    1995-11-01

    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

  6. WWER NPPs fuel handling machine control system

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  7. Thermal and mechanical analyses of the spent nuclear fuel disposal canister and its barriers according to the design variable change

    International Nuclear Information System (INIS)

    Kwon, Young Joo

    2006-03-01

    -plastic materials. In this report, additionally, the nonlinear structural analysis of the composite system, i.e., 'canister + buffer' structural system, due to the 10cm sudden rock movement caused by earthquake etc., is conducted to determine the structural safety of the KDC-1 canister model with diameter of D=102cm. Summarizing all the structural analysis results of the canister, the structural safety of the KDC-1 canister model with diameter with D=102cm is secured for the falling accidents happened while handling the canister in the repository and for the 10cm sudden rock movement cause by the earthquake etc. The rigid body dynamic analysis of the canister for the falling accident is carried out using the CAE system, RecurDyn, and the nonlinear structural analysis is performed using the finite element analysis code, NISA

  8. End of FY2014 Report - Filter Measurement System for Nuclear Material Storage Canisters (Including Altitude Correction for Filter Pressure Drop)

    Energy Technology Data Exchange (ETDEWEB)

    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)

    2015-02-24

    Two LANL FTS (Filter Test System ) devices for nuclear material storage canisters are fully operational. One is located in PF-4 ( i.e. the TA-55 FTS) while the other is located at the Radiation Protection Division’s Aerosol Engineering Facility ( i.e. the TA-3 FTS). The systems are functionally equivalent , with the TA-3 FTS being the test-bed for new additions and for resolving any issues found in the TA-55 FTS. There is currently one unresolved issue regarding the TA-55 FTS device. The canister lid clamp does not give a leak tight seal when testing the 1 QT (quart) or 2 QT SAVY lids. An adapter plate is being developed that will ensure a correct test configuration when the 1 or 2 QT SAVY lid s are being tested .

  9. CFD Analysis on the Passive Heat Removal by Helium and Air in the Canister of Spent Fuel Dry Storage System

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Do Young; Jeong, Ui Ju; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2016-05-15

    In the current commercial design, the canister of the dry storage system is mainly backfilled with helium gas. Helium gas shows very conductive behavior due to high thermal conductivity and small density change with temperature. However, other gases such as air, argon, or nitrogen are expected to show effective convective behavior. Thus these are also considered as candidates for the backfill gas to provide effective coolability. In this study, to compare the dominant cooling mechanism and effectiveness of cooling between helium gas and air, a computational fluid dynamics (CFD) analysis for the canister of spent fuel dry storage system with backfill gas of helium and air is carried out. In this study, CFD simulations for the helium and air backfilled gas for dry storage system canister were carried out using ANSYS FLUENT code. For the comparison work, two backfilled fluids were modeled with same initial and boundary conditions. The observed major difference can be summarized as follows. - The simulation results showed the difference in dominant heat removal mechanism. Conduction for helium, and convection for air considering Reynolds number distribution. - The temperature gradient inside the fuel assembly showed that in case of air, more effective heat mixing occurred compared to helium.

  10. 30 CFR 75.817 - Cable handling and support systems.

    Science.gov (United States)

    2010-07-01

    ... High-Voltage Longwalls § 75.817 Cable handling and support systems. Longwall mining equipment must be provided with cable-handling and support systems that are constructed, installed and maintained to minimize... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Cable handling and support systems. 75.817...

  11. The role of the canister in a system for the final disposal of spent fuel or high-level waste

    International Nuclear Information System (INIS)

    Papp, T.

    1986-01-01

    A final repository for radioactive waste must isolate the toxic substances or distribute their release over time or space to avoid causing harmful concentrations of radionuclides in the biosphere. The Swedish research has focused on a repository 500 m down in crystalline rock where the geochemical environment can give canisters a service life of the order of a million years. These evaluations are discussed and the safety effect of the canister is compared with that of other barriers available in a repository system. Our conclusions are that a combined protection effect of natural and man-made barriers can be achieved that substantially exceeds what could reasonably be required by society. An actual repository design can then be based on an optimization of the cost to reach a level of accepted safety with due regard for the safety margins and redundancy necessary for achieving public confidence. (author)

  12. 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.

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

    International Nuclear Information System (INIS)

    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

  14. 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.

  15. 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...

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

    International Nuclear Information System (INIS)

    Henttonen, V.; Suikki, M.

    1992-08-01

    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

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

    International Nuclear Information System (INIS)

    Henttonen, V.; Suikki, M.

    1992-06-01

    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)

  18. Aerobot Sampling and Handling System, Phase I

    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...

  19. TRADITIONAL CANISTER-BASED OPEN WASTE MANAGEMENT SYSTEM VERSUS CLOSED SYSTEM: HAZARDOUS EXPOSURE PREVENTION AND OPERATING THEATRE STAFF SATISFACTION.

    Science.gov (United States)

    Horn, M; Patel, N; MacLellan, D M; Millard, N

    2016-06-01

    Exposure to blood and body fluids is a major concern to health care professionals working in operating rooms (ORs). Thus, it is essential that hospitals use fluid waste management systems that minimise risk to staff, while maximising efficiency. The current study compared the utility of a 'closed' system with a traditional canister-based 'open' system in the OR in a private hospital setting. A total of 30 arthroscopy, urology, and orthopaedic cases were observed. The closed system was used in five, four, and six cases, respectively and the open system was used in nine, two, and four cases, respectively. The average number of opportunities for staff to be exposed to hazardous fluids were fewer for the closed system when compared to the open during arthroscopy and urology procedures. The open system required nearly 3.5 times as much staff time for set-up, maintenance during procedures, and post-procedure disposal of waste. Theatre staff expressed greater satisfaction with the closed system than with the open. In conclusion, compared with the open system, the closed system offers a less hazardous and more efficient method of disposing of fluid waste generated in the OR.

  20. WASTE HANDLING BUILDING FIRE PROTECTION SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bigbee

    2000-06-21

    The Waste Handling Building Fire Protection System provides the capability to detect, control, and extinguish fires and/or mitigate explosions throughout the Waste Handling Building (WHB). Fire protection includes appropriate water-based and non-water-based suppression, as appropriate, and includes the distribution and delivery systems for the fire suppression agents. The Waste Handling Building Fire Protection System includes fire or explosion detection panel(s) controlling various detectors, system actuation, annunciators, equipment controls, and signal outputs. The system interfaces with the Waste Handling Building System for mounting of fire protection equipment and components, location of fire suppression equipment, suppression agent runoff, and locating fire rated barriers. The system interfaces with the Waste Handling Building System for adequate drainage and removal capabilities of liquid runoff resulting from fire protection discharges. The system interfaces with the Waste Handling Building Electrical Distribution System for power to operate, and with the Site Fire Protection System for fire protection water supply to automatic sprinklers, standpipes, and hose stations. The system interfaces with the Site Fire Protection System for fire signal transmission outside the WHB as needed to respond to a fire emergency, and with the Waste Handling Building Ventilation System to detect smoke and fire in specific areas, to protect building high-efficiency particulate air (HEPA) filters, and to control portions of the Waste Handling Building Ventilation System for smoke management and manual override capability. The system interfaces with the Monitored Geologic Repository (MGR) Operations Monitoring and Control System for annunciation, and condition status.

  1. WASTE HANDLING BUILDING FIRE PROTECTION SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    J. D. Bigbee

    2000-01-01

    The Waste Handling Building Fire Protection System provides the capability to detect, control, and extinguish fires and/or mitigate explosions throughout the Waste Handling Building (WHB). Fire protection includes appropriate water-based and non-water-based suppression, as appropriate, and includes the distribution and delivery systems for the fire suppression agents. The Waste Handling Building Fire Protection System includes fire or explosion detection panel(s) controlling various detectors, system actuation, annunciators, equipment controls, and signal outputs. The system interfaces with the Waste Handling Building System for mounting of fire protection equipment and components, location of fire suppression equipment, suppression agent runoff, and locating fire rated barriers. The system interfaces with the Waste Handling Building System for adequate drainage and removal capabilities of liquid runoff resulting from fire protection discharges. The system interfaces with the Waste Handling Building Electrical Distribution System for power to operate, and with the Site Fire Protection System for fire protection water supply to automatic sprinklers, standpipes, and hose stations. The system interfaces with the Site Fire Protection System for fire signal transmission outside the WHB as needed to respond to a fire emergency, and with the Waste Handling Building Ventilation System to detect smoke and fire in specific areas, to protect building high-efficiency particulate air (HEPA) filters, and to control portions of the Waste Handling Building Ventilation System for smoke management and manual override capability. The system interfaces with the Monitored Geologic Repository (MGR) Operations Monitoring and Control System for annunciation, and condition status

  2. Structural analysis of fuel handling systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  3. Structural analysis of fuel handling systems

    International Nuclear Information System (INIS)

    Lee, L.S.S.

    1996-01-01

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

  4. Conceptual designs of radioactive canister transporters

    International Nuclear Information System (INIS)

    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

  5. Acceptance for Beneficial Use for the Canister Cleaning System for the K West basin Project - A.2.A

    International Nuclear Information System (INIS)

    FARWICK, C.C.

    2000-01-01

    This documents the documentation that is required to be turned over to Operations with the Canister Cleaning System (CCS). The Acceptance for Beneficial Use will be updated as required prior to turnover. This document is prepared for the purposes of documenting an agreement among the various disciplines and organizations within the Spent Nuclear Fuel (SNF) Project as to what is required in terms of installed components of the CCS. This documentation will be used to achieve project closeout and turnover of ownership of the CCS to K Basins Operations

  6. 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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1996-11-01

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

  9. Analytical Evaluation of Preliminary Drop Tests Performed to Develop a Robust Design for the Standardized DOE Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    Ware, A.G.; Morton, D.K.; Smith, N.L.; Snow, S.D.; Rahl, T.E.

    1999-01-01

    The Department of Energy (DOE) has developed a design concept for a set of standard canisters for the handling, interim storage, transportation, and disposal in the national repository, of DOE spent nuclear fuel (SNF). The standardized DOE SNF canister has to be capable of handling virtually all of the DOE SNF in a variety of potential storage and transportation systems. It must also be acceptable to the repository, based on current and anticipated future requirements. This expected usage mandates a robust design. The canister design has four unique geometries, with lengths of approximately 10 feet or 15 feet, and an outside nominal diameter of 18 inches or 24 inches. The canister has been developed to withstand a drop from 30 feet onto a rigid (flat) surface, sustaining only minor damage - but no rupture - to the pressure (containment) boundary. The majority of the end drop-induced damage is confined to the skirt and lifting/stiffening ring components, which can be removed if de sired after an accidental drop. A canister, with its skirt and stiffening ring removed after an accidental drop, can continue to be used in service with appropriate operational steps being taken. Features of the design concept have been proven through drop testing and finite element analyses of smaller test specimens. Finite element analyses also validated the canister design for drops onto a rigid (flat) surface for a variety of canister orientations at impact, from vertical to 45 degrees off vertical. Actual 30-foot drop testing has also been performed to verify the final design, though limited to just two full-scale test canister drops. In each case, the analytical models accurately predicted the canister response

  10. Neutron interrogator assay system for the Idaho Chemical Processing Plant waste canisters and spent fuel: preliminary description and operating procedures manual

    International Nuclear Information System (INIS)

    Menlove, H.O.; Eccleston, G.; Close, D.A.; Speir, L.G.

    1978-05-01

    A neutron interrogation assay system is being designed for the measurement of waste canisters and spent fuel packages at the new Idaho Chemical Processing Plant to be operated by Allied Chemical Corp. The assay samples consist of both waste canisters from the fluorinel dissolution process and spent fuel assemblies. The assay system is a 252 Cf ''Shuffler'' that employs a cyclic sequence of fast-neutron interrogation with a 252 Cf source followed by delayed-neutron counting to determine the 235 U content

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

    International Nuclear Information System (INIS)

    Butler, J.W.; Hodges, D.; Royston, R.

    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

  12. Data-handling system for the Fly's Eye experiment

    International Nuclear Information System (INIS)

    Bergeson, H.E.; Cassiday, G.L.; Cooper, D.A.

    1975-01-01

    The Fly's Eye air scintillation experiment presents severe data-handling requirements for two reasons. First, nearly 1,000 photomultipliers each produce outputs at rates from 100 Khz to 20 Mhz. Second, much of the signal arrives before a trigger is formed. A data handling system which will deal with this problem is described. (orig.) [de

  13. Handling effluent from nuclear thermal propulsion system ground tests

    International Nuclear Information System (INIS)

    Shipers, L.R.; Allen, G.C.

    1992-01-01

    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

  14. Cable handling system for use in a nuclear reactor

    International Nuclear Information System (INIS)

    Crosgrove, R.O.; Larson, E.M.; Moody, E.

    1982-01-01

    A cable handling system for use in an installation such as a nuclear reactor is disclosed herein along with relevant portions of the reactor which, in a preferred embodiment, is a liquid metal fast breeder reactor. The cable handling system provides a specific way of interconnecting certain internal reactor components with certain external components, through an assembly of rotatable plugs. Moreover, this is done without having to disconnect these components from one another during rotation of the plugs and yet without interfering with other reactor components in the vicinity of the rotating plugs and cable handling system

  15. Advanced handling-systems with enhanced performance flexibility

    International Nuclear Information System (INIS)

    1986-04-01

    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) [de

  16. 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.

  17. System and method for slurry handling

    Science.gov (United States)

    Steele, Raymond Douglas; Oppenheim, Judith Pauline

    2015-12-29

    A system includes a slurry depressurizing system that includes a liquid expansion system configured to continuously receive a slurry at a first pressure and continuously discharge the slurry at a second pressure. For example, the slurry depressurizing system may include an expansion turbine to expand the slurry from the first pressure to the second pressure.

  18. Effect of Canister Movement on Water Turbidity

    International Nuclear Information System (INIS)

    TRIMBLE, D.J.

    2000-01-01

    Requirements for evaluating the adherence characteristics of sludge on the fuel stored in the K East Basin and the effect of canister movement on basin water turbidity are documented in Briggs (1996). The results of the sludge adherence testing have been documented (Bergmann 1996). This report documents the results of the canister movement tests. The purpose of the canister movement tests was to characterize water turbidity under controlled canister movements (Briggs 1996). The tests were designed to evaluate methods for minimizing the plumes and controlling water turbidity during fuel movements leading to multi-canister overpack (MCO) loading. It was expected that the test data would provide qualitative visual information for use in the design of the fuel retrieval and water treatment systems. Video recordings of the tests were to be the only information collected

  19. Overhead remote handling systems for the process facility modifications project

    International Nuclear Information System (INIS)

    Wiesener, R.W.; Grover, D.L.

    1987-01-01

    Each of the cells in the process facility modifications (PFM) project complex is provided with a variety of general purpose remote handling equipment including bridge cranes, monorail hoist, bridge-mounted electromechanical manipulator (EMM) and an overhead robot used for high efficiency particulate air (HEPA) filter changeout. This equipment supplements master-slave manipulators (MSMs) located throughout the complex to provide an overall remote handling system capability. The overhead handling equipment is used for fuel and waste material handling operations throughout the process cells. The system also provides the capability for remote replacement of all in-cell process equipment which may fail or be replaced for upgrading during the lifetime of the facility

  20. Ontario Hydro Pickering Generating Station fuel handling system performance

    International Nuclear Information System (INIS)

    Underhill, H.J.

    1986-01-01

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

  1. Conceptual design report for a remotely operated cask handling system

    International Nuclear Information System (INIS)

    Yount, J.A.; Berger, J.D.

    Recent advances in remote handling utilizing commercial robotics are conceptually applied to the problem of lowering operator cumulative dose and increasing throughput during cask handling operations in proposed nuclear waste container shipping and receiving facilities. The functional criteria for each subsystem are defined, and candidate systems are described. The report also contains a generic description of a waste receiving facility, to show possible deployment configurations for the equipment

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

    International Nuclear Information System (INIS)

    Ahmed, H.; Cowie, A.; Ma, D.

    1978-10-01

    SLSF loop handling system was analyzed for deadweight and postulated dynamic loading conditions, identified in Chapters II and III in Volume I of this report, using a linear elastic static equivalent method of stress analysis. Stress analysis of the loop handling machine is presented in Volume I of this report. Chapter VII in Volume I of this report is a contribution by EG and G Co., who performed the work under ANL supervision

  3. SRL canister impact tests

    International Nuclear Information System (INIS)

    Kelker, J.W. Jr.

    1986-05-01

    The Defense Waste Processing Facility (DWPF) is being constructed at the SRP for the containerization of high-level nuclear waste as a waste form for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in Type 304L stainless steel canisters 2 feet in diameter x 9 feet 10 inches long. The canisters have a minimum wall thickness of 3/8 inch. Over a three-year period, nineteen drop-tests of nine canisters, filled with simulated waste glass, were made in support of the DWPF containerization program. Eight of the canister evaluation tests were of Type 304L stainless steel material and one was of commercially pure titanium. Three different length (9.44, 5.06, and 7.88 inch) nozzle configurations containing final closure upset welds were evaluated for the stainless steel canisters. All impact tests of the stainless steel canisters, which included bottom-, side-, and top-drops, were acceptable. The bottom-drop test of the titanium canister, which contained a final closure upset weld, was acceptable; however, the top-drop resulted in a breaching of the top head where it joins the nozzle. The final closure titanium upset weld was acceptable. The titanium canister wall thickness was 1/4 inch

  4. A graphics based remote handling control system

    International Nuclear Information System (INIS)

    Leinemann, K.

    1984-08-01

    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.) [de

  5. Computer aided accountancy for tritium handling systems

    International Nuclear Information System (INIS)

    Spannagel, G.; Schmid, C.

    1993-03-01

    A program is presented which can be used in tritium bearing systems for inventory taking and accountancy purposes. In particular, a detailed description is given of the environment in which this program has been integrated. It is explained in which way a high user friendliness has been attained and which structures contribute to achieving the flexibility required. (orig.) [de

  6. 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.

  7. Mechanical integrity of canisters

    International Nuclear Information System (INIS)

    Nilsson, Fred

    1992-12-01

    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

  8. The concrete canister program

    International Nuclear Information System (INIS)

    Ohta, M.M.

    1978-02-01

    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)

  9. Fuel handling system of nuclear reactor plants

    International Nuclear Information System (INIS)

    Faulstich, D.L.

    1991-01-01

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

  10. Evolution of the Darlington NGS fuel handling computer systems

    International Nuclear Information System (INIS)

    Leung, V.; Crouse, B.

    1996-01-01

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

  11. Evolution of the Darlington NGS fuel handling computer systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  12. Remote systems and automation in radioactive waste package handling

    International Nuclear Information System (INIS)

    Gneiting, B.C.; Hayward, M.L.

    1987-01-01

    A proof-of-principle test was conducted at the Hanford Engineering Development Laboratory (HEDL) to demonstrate the feasibility of performing cask receiving and unloading operations in a remote and partially automated manner. This development testing showed feasibility of performing critical cask receipt, preparation, and unloading operations from a single control station using remote controls and indirect viewing. Using robotics and remote automation in a cask handling system can result in lower personnel exposure levels and cask turnaround times while maintaining operational flexibility. An automated cask handling system presents a flexible state-of-the-art, cost effective alternative solution to hands-on methods that have been used in the past

  13. Safety handling manual for high dose rate remote afterloading system

    International Nuclear Information System (INIS)

    1999-01-01

    This manual is mainly for safety handling of 192 Ir-RALS (remote afterloading system) of high dose rate and followings were presented: Procedure and document format for the RALS therapy and for handling of its radiation source with the purpose of prevention of human errors and unexpected accidents, Procedure for preventing errors occurring in the treatment schedule and operation, and Procedure and format necessary for newly introducing the system into a facility. Consistency was intended in the description with the quality assurance guideline for therapy with small sealed radiation sources made by JASTRO (Japan Society for Therapeutic Radiology and Oncology). Use of the old type 60 Co-RALS was pointed out to be a serious problem remained and its safety handling procedure was also presented. (K.H.)

  14. 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

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

    International Nuclear Information System (INIS)

    Robinson, Peter

    2006-03-01

    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

  16. Conceptual design of divertor cassette handling by remote handling system for JT-60SA

    International Nuclear Information System (INIS)

    Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

    2007-01-01

    The JT-60SA aims to contribute and supplement ITER toward DEMO reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is prohibited. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor modules. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor module, which is 10 degrees wide in toroidal direction and weighs 500kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor module to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the module by a pallet installed from outside the VV. (author)

  17. Conceptual design of divertor cassette handling by remote handling system of JT-60SA

    International Nuclear Information System (INIS)

    Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

    2008-01-01

    The JT-60SA aims to contribute and supplement ITER toward demonstration fusion reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is restricted. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor cassettes. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor cassette, which is 10 degrees wide in toroidal direction and weighs 500 kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor cassette to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the cassette by a pallet installed from outside the VV. (author)

  18. The prospective usage of the multi-purpose canister and impacts on the waste management and disposal system

    International Nuclear Information System (INIS)

    McLeod, N.B.

    1993-01-01

    The Multi-Purpose Canister (MPC) is designed to be loaded with spent fuel and sealed at reactors and then serve the functions of transport, storage and disposal without reopening. It can be either self-shielded or unshielded, thus requiring compatible overpacks for transport, storage and disposal. The MPC is not a new concept but it may now be viable because of the particular characteristics at Yucca Mountain: larger MPCs are possible because of ramp access to the repository horizon, and the less difficult temperature limits because of in-drift emplacement, rather than borehole emplacement. This paper describes the advantages and disadvantages of adopting the MPC as the principal technology to be employed in the US program. Use of the MPC permits integration of the utility and DOE portions of the system as well as among the elements within the DOE portion. Paradoxically, the principal disadvantage of the MPC is a direct consequence of its merit as an integrating technology. Full integration includes disposability without reopening, and requires that disposability design decisions be made and implemented well in advance of when waste package licensing uncertainties are resolved. There is, therefore, a risk that MPCs loaded prior to waste package licensing will have to be opened. This risk is discussed in terms of probability and consequences and various alternatives for mitigating this risk are discussed

  19. Modelling dust liberation in bulk material handling systems

    NARCIS (Netherlands)

    Derakhshani, S.M.

    2016-01-01

    Dust has negative effects on the environmental conditions, human health as well as industrial equipment and processes. In this thesis, the transfer point of a belt conveyor as a bulk material handling system with a very high potential place for dust liberation is studied. This study is conducted

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

    International Nuclear Information System (INIS)

    TEDESCHI, D.J.

    2000-01-01

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

  1. Advanced robotic remote handling system for reactor dismantlement

    International Nuclear Information System (INIS)

    Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

    1991-01-01

    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)

  2. Bifurcation methods of dynamical systems for handling nonlinear ...

    Indian Academy of Sciences (India)

    physics pp. 863–868. Bifurcation methods of dynamical systems for handling nonlinear wave equations. DAHE FENG and JIBIN LI. Center for Nonlinear Science Studies, School of Science, Kunming University of Science and Technology .... (b) It can be shown from (15) and (18) that the balance between the weak nonlinear.

  3. Superphenix 1 primary handling system fabrication and testing

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements

    International Nuclear Information System (INIS)

    KLEM, M.J.

    2000-01-01

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the

  5. Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

    International Nuclear Information System (INIS)

    Lee, J.Y.; Cho, D.K.; Lee, M.S.; Kook, D.H.; Choi, H.J.; Choi, J.W.; Wang, L.M.

    2012-01-01

    Highlights: ► A reference disposal concept for spent nuclear fuels in Korea has been reviewed. ► To enhance the disposal efficiency, alternative disposal concepts were developed. ► Thermal analyses for alternative disposal concepts were performed. ► From the result of the analyses, the disposal efficiency of the concepts was reviewed. ► The most effective concept was suggested. - Abstract: Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.

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

    International Nuclear Information System (INIS)

    Sreten Mastilovic

    2001-01-01

    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

  7. Protecting worker health and safety using remote handling systems

    International Nuclear Information System (INIS)

    Dennison, D.K.; Merrill, R.D.; Reed, R.K.

    1995-03-01

    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

  8. Feasibility study of CANDU-9 fuel handling system

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  9. A smartphone controlled handheld microfluidic liquid handling system.

    Science.gov (United States)

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

    2014-10-21

    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 HIV1 p24 sandwich immunoassay on a multi-layer PDMS chip without any human intervention. The footprint of the system is 6 × 10.5 × 16.5 cm, and the total weight is 829 g including battery. Powered by a 12.8 V 1500 mAh Li battery, the system consumed 2.2 W on average during the immunoassay and lasted for 8.7 h. This handheld microfluidic liquid handling platform is generally applicable to many biochemical and cell-based assays requiring complex liquid manipulation and sample preparation steps such as FISH, PCR, flow cytometry and nucleic acid sequencing. In particular, the integration of this technology with read-out biosensors may help enable the realization of the long-sought Tricorder-like handheld in vitro diagnostic (IVD) systems.

  10. Mockup of an automated material transport system for remote handling

    International Nuclear Information System (INIS)

    Porter, M.L.

    1992-01-01

    An Automated Material Transport System (AMTS) was identified for transport of samples within a Material and Process Control Laboratory (MPCL). The MPCL was designed with a dry sample handling laboratory and a wet chemistry analysis laboratory. Each laboratory contained several processing gloveboxes. The function of the AMTS was to automate the handling of materials, multiple process samples, and bulky items between process stations with a minimum of operator intervention and with minimum o[ waiting periods and nonproductive activities. This paper discusses the system design features, capabilities and results of initial testing. The overall performance of the AMTS is very good. No major problems or concerns were identified. System commands are simple and logical making the system user friendly. Operating principle and design of individual components is simple. With the addition of various track modules, the system can be configured in most any configuration. The AMTS lends itself very well for integration with other automated systems or products. The AMTS is suited for applications involving light payloads which require multiple sample and material handling, lot tracking, and system integration with other products

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

    International Nuclear Information System (INIS)

    Tamaoki, Tetsuo; Yamamoto, Hiroki; Sato, Masuo; Yoshida, Megumu; Kaneko, Tomoko; Terunuma, Seiichi; Takatsuto, Hiroshi; Morimoto, Makoto.

    1992-01-01

    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)

  12. CLASSIFICATION OF THE MGR WASTE HANDLING BUILDING ELECTRICAL SYSTEM

    International Nuclear Information System (INIS)

    S.E. Salzman

    1999-01-01

    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)

  13. Contamination confinement system of irradiated materials handling laboratories

    International Nuclear Information System (INIS)

    Lobao, A. dos S.T.; Araujo, J.A. de; Camilo, R.L.

    1988-06-01

    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) [pt

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

    International Nuclear Information System (INIS)

    Merrill, R.D.; Hurd, R.; Couture, S.; Wilhelmsen, K.

    1995-02-01

    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. Containment system of contamination in irradiated materials handling laboratories

    International Nuclear Information System (INIS)

    Lobao, A.S.T.; Araujo, J.A. de; Camilo, R.L.

    1988-01-01

    A study to prevent radiactivity 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 specification of the working areas, ducts and filtering systems in order to get the best conditions for the safe handling of irradiated materials. (author) [pt

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

    International Nuclear Information System (INIS)

    Shapiro, S.A.; Story, G.L.

    1979-01-01

    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

  17. Manufacture of disposal canisters

    International Nuclear Information System (INIS)

    Nolvi, L.

    2009-12-01

    The report summarizes the development work carried out in the manufacturing of disposal canister components, and present status, in readiness for manufacturing, of the components for use in assembly of spent nuclear fuel disposal canister. The disposal canister consist of two major components: the nodular graphite cast iron insert and overpack of oxygen-free copper. The manufacturing process for copper components begins with a cylindrical cast copper billet. Three different manufacturing processes i.e. pierce and draw, extrusion and forging are being developed, which produce a seamless copper tube or a tube with an integrated bottom. The pierce and draw process, Posiva's reference method, makes an integrated bottom possible and only the lid requires welding. Inserts for BWR-element are cast with 12 square channels and inserts for VVER 440-element with 12 round channels. Inserts for EPR-elements have four square channels. Casting of BWR insert type has been studied so far. Experience of casting inserts for PWR, which is similar to the EPR-type, has been got in co-operation with SKB. The report describes the processes being developed for manufacture of disposal canister components and some results of the manufacturing experiments are presented. Quality assurance and quality control in manufacture of canister component is described. (orig.)

  18. DOE requests waiver on double containment for HLW canisters

    International Nuclear Information System (INIS)

    Lobsenz, G.

    1994-01-01

    The Energy Department has asked the Nuclear Regulatory Commission to waive double containment requirements for vitrified high-level radioactive waste canisters, saying the additional protection is not necessary and too costly. NRC said it had received a petition from DOE contending that the vitrified waste canisters were durable enough without double containment to prevent any potential plutonium release during handling and shipping. DOE said testing had shown that the vitrified waste canisters were similar - even superior - in durability to spent reactor fuel shipments, which NRC specifically exempted from the double containment requirement

  19. 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

  20. SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    S. C. Khamankar

    2000-01-01

    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. Remote systems and automation in radioactive waste package handling

    International Nuclear Information System (INIS)

    Gneiting, B.C.; Hayward, M.L.

    1987-01-01

    A proof-of-principle test was conducted at the Hanford Engineering Development Laboratory (HEDL) to demonstrate the feasibility of performing cask receiving and unloading operations in a remote and partially automated manner. This development testing showed feasibility of performing critical cask receipt, preparation, and unloading operations from a single control station using remote controls and indirect viewing. Using robotics and remote automation in a cask handling system can result in lower personnel exposure levels and cask turnaround times while maintaining operational flexibility. An automated cask handling system presents a flexible state-of-the-art, cost effective alternative solution to hands-on methods that have been used in the past. 7 refs., 13 figs

  2. Development and implementation of automated radioactive materials handling systems

    International Nuclear Information System (INIS)

    Jacoboski, D.L.

    1992-12-01

    Material handling of radioactive and hazardous materials has forced the need to pursue remotely operated and robotic systems in light of operational safety concerns. Manual maneuvering, repackaging, overpacking and inspecting of containers which store radioactive and hazardous materials is the present mode of operation at the Department of Energy (DOE) Fernald Environmental Management Project (FEMP) in Fernald Ohio. The manual methods are unacceptable in the eyes of concerned site workers and influential community oversight committees. As an example to respond to the FEMP material handling needs, design efforts have been initiated to provide a remotely operated system to repackage thousands of degradated drums containing radioactive Thorium: Later, the repackaged Thorium will be shipped offsite to a predesignated repository again requiring remote operation

  3. Mockup of an automated material transport system for remote handling

    International Nuclear Information System (INIS)

    Porter, M.L.

    1992-01-01

    The automated material transport system (AMTS) was conceived for the transport of samples within the material and process control laboratory (MPCL), located in the plutonium processing building of the special isotope separation (SIS) facility. The MPCL was designed with a dry sample handling laboratory and a wet chemistry analysis laboratory. Each laboratory contained several processing glove boxes. The function of the AMTS was to automate the handling of materials, multiple process samples, and bulky items between process stations with a minimum of operator intervention and with a minimum of waiting periods and nonproductive activities. The AMTS design requirements, design verification mockup plan, and AMTS mockup procurement specification were established prior to cancellation of the SIS project. Due to the AMTS's flexibility, the need for technology development, and applicability to other US Department of Energy facilities, mockup of the AMTS continued. This paper discusses the system design features, capabilities, and results of initial testing

  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)

    Drotning, W.; Kimberly, H.; Wapman, W.; Darras, D.

    1997-01-01

    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. A computerized system for handling renal size measurements from urograms

    International Nuclear Information System (INIS)

    Claesson, I.; Jacobsson, B.F.; Riha, M.

    1987-01-01

    The size of a kidney, as measured on an urogram, is a sensitive indicator of renal damage in a child with urinary tract infection, and renal surface area correlates well with glomerular filtration rate. Sequential measurements can be invaluable in evaluating the efficacy of a regimen of treatment. A system utilizing a personal microcomputer has been developed to facilitate the measuring procedure and the handling and analysis of data. (orig.)

  7. System design for safe robotic handling of nuclear materials

    International Nuclear Information System (INIS)

    Drotning, W.; Wapman, W.; Fahrenholtz, J.; Kimberly, H.; Kuhlmann, J.

    1996-01-01

    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. Engineered barrier systems and canister orientation studies for the Yucca Mountain Project, Nevada

    International Nuclear Information System (INIS)

    Wilder, D.G.

    1990-07-01

    Emplacement borehole orientation directly impacts many aspects of the Engineered Barrier System (EBS) and interactions with the near field environment. This paper considers the impacts of orientation on the hydrologic portion of the environment and its interactions with the EBS. The hydrologic environments is considered from a conceptual standpoint, the numerical analyses are left for subsequent work. As reported in this paper, several aspects of the hydrological environment are more favorable for long term performance of vertically oriented rather than horizontally oriented Waste Packages. 19 refs., 15 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    TU, K.C.

    1999-10-08

    Multi-Canister Overpacks (MCOs) containing spent nuclear fuel (SNF) will be routinely handled at the Canister Storage Building (CSB) during fuel movement operations in the SNF Project. This analysis was performed to investigate the potential for damage from an eccentric accidental drop onto the standard storage tube, overpack tube, service station, or sample/weld station. Appendix D was added to the FDNW document to include the peer Review Comment Record & transmittal record.

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

    International Nuclear Information System (INIS)

    TU, K.C.

    1999-01-01

    Multi-Canister Overpacks (MCOs) containing spent nuclear fuel (SNF) will be routinely handled at the Canister Storage Building (CSB) during fuel movement operations in the SNF Project. This analysis was performed to investigate the potential for damage from an eccentric accidental drop onto the standard storage tube, overpack tube, service station, or sample/weld station. Appendix D was added to the FDNW document to include the peer Review Comment Record and transmittal record

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

    International Nuclear Information System (INIS)

    Lopez, D.A.; Abbott, D.G.

    1994-02-01

    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

  12. Clay Generic Disposal System Model - Sensitivity Analysis for 32 PWR Assembly Canisters (+2 associated model files).

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Edgar [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-10-01

    The Used Fuel Disposition Campaign (UFDC), as part of the DOE Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technology program (FCT) is investigating the disposal of high level radioactive waste (HLW) and spent nuclear fuela (SNF) in a variety of geologic media. The feasibility of disposing SNF and HLW in clay media has been investigated and has been shown to be promising [Ref. 1]. In addition the disposal of these wastes in clay media is being investigated in Belgium, France, and Switzerland. Thus, Argillaceous media is one of the environments being considered by UFDC. As identified by researchers at Sandia National Laboratory, potentially suitable formations that may exist in the U.S. include mudstone, clay, shale, and argillite formations [Ref. 1]. These formations encompass a broad range of material properties. In this report, reference to clay media is intended to cover the full range of material properties. This report presents the status of the development of a simulation model for evaluating the performance of generic clay media. The clay Generic Disposal System Model (GDSM) repository performance simulation tool has been developed with the flexibility to evaluate not only different properties, but different waste streams/forms and different repository designs and engineered barrier configurations/ materials that could be used to dispose of these wastes.

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

    International Nuclear Information System (INIS)

    Arapakota, D.

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

  15. Availability analysis of the ITER blanket remote handling system

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  16. Three-dimensional television system for remote handling

    International Nuclear Information System (INIS)

    Dumbreck, A.A.; Abel, E.

    1988-01-01

    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)

  17. Systems for harvesting and handling cotton plant residue

    Energy Technology Data Exchange (ETDEWEB)

    Coates, W. [Univ. of Arizona, Tucson, AZ (United States)

    1993-12-31

    In the warmer regions of the United States, cotton plant residue must be buried to prevent it from serving as an overwintering site for insect pests such as the pink bollworm. Most of the field operations used to bury the residue are high energy consumers and tend to degrade soil structure, thereby increasing the potential for erosion. The residue is of little value as a soil amendment and consequently is considered a negative value biomass. A commercial system to harvest cotton plant residue would be of both economic and environmental benefit to cotton producers. Research has been underway at the University of Arizona since the spring of 1991 to develop a commercially viable system for harvesting cotton plant residue. Equipment durability, degree of densification, energy required, cleanliness of the harvested material, and ease of product handling and transport are some of the performance variables which have been measured. Two systems have proven superior. In both, the plants are pulled from the ground using an implement developed specifically for the purpose. In one system, the stalks are baled using a large round baler, while in the other the stalks are chopped with a forage harvester, and then made into packages using a cotton module maker. Field capacities, energy requirements, package density and durability, and ease of handling with commercially available equipment have been measured for both systems. Selection of an optimum system for a specific operation depends upon end use of the product, and upon equipment availability.

  18. A Film Canister Colorimeter.

    Science.gov (United States)

    Gordon, James; James, Alan; Harman, Stephanie; Weiss, Kristen

    2002-01-01

    A low-cost, low-tech colorimeter was constructed from a film canister. The student-constructed colorimeter was used to show the Beer-Lambert relationship between absorbance and concentration and to calculate the value of the molar absorptivity for permanganate at the wavelength emission maximum for an LED. Makes comparisons between this instrument…

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

    Energy Technology Data Exchange (ETDEWEB)

    Bye, A; Moum, B R [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-09-01

    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.

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

    International Nuclear Information System (INIS)

    Bye, A.; Moum, B.R.

    1997-01-01

    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

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

    International Nuclear Information System (INIS)

    Romesberg, L.E.; Pope, R.B.; Burgoyne, R.M.

    1982-04-01

    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

  2. Handling encapsulated spent fuel in a geologic repository environment

    International Nuclear Information System (INIS)

    Ballou, L.B.

    1983-02-01

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

  3. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    Kukkola, T.; Toermaelae, V.P.

    2005-06-01

    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 the 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 100m/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. (orig.)

  4. Progress in standardization for ITER Remote Handling control system

    International Nuclear Information System (INIS)

    Hamilton, David Thomas; Tesini, Alessandro; Ranz, Roberto; Kozaka, Hiroshi

    2014-01-01

    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

  5. Progress in standardization for ITER Remote Handling control system

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, David Thomas, E-mail: david.hamilton@iter.org [ITER Organization, Route de Vinon, 13115 St. Paul-lez-Durance (France); Tesini, Alessandro [ITER Organization, Route de Vinon, 13115 St. Paul-lez-Durance (France); Ranz, Roberto [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Kozaka, Hiroshi [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan)

    2014-10-15

    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.

  6. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1993-01-01

    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

  7. MCC-15: waste/canister accident testing and analysis method

    International Nuclear Information System (INIS)

    Slate, S.C.; Pulsipher, B.A.; Scott, P.A.

    1985-02-01

    The Materials Characterization Center (MCC) at the Pacific Northwest Laboratory (PNL) is developing standard tests to characterize the performance of nuclear waste forms under normal and accident conditions. As part of this effort, the MCC is developing MCC-15, Waste/Canister Accident Testing and Analysis. MCC-15 is used to test canisters containing simulated waste forms to provide data on the effects of accidental impacts on the waste form particle size and on canister integrity. The data is used to support the design of transportation and handling equipment and to demonstrate compliance with repository waste acceptance specifications. This paper reviews the requirements that led to the development of MCC-15, describes the test method itself, and presents some early results from tests on canisters representative of those proposed for the Defense Waste Processing Facility (DWPF). 13 references, 6 figures

  8. TMI-2 fuel canister interface requirements for INEL. Revision 1

    International Nuclear Information System (INIS)

    Wilkins, D.E.; Martz, D.E.; Reno, H.W.

    1984-06-01

    This report focuses on fuel canister interface requirements at INEL which should be incorporated into the canister design criteria. The requirements will ensure compatibility with existing INEL structures and equipment to be used for receipt, unloading, and storage of fuel canisters. INEL can and does receive and store radioactive materials in many different forms, including reactor fuel. INEL requires detailed descriptions of canisters and casks. Therefore, requirements listed represent engineering design features which will simplify the handling and storage operations; consequently, they are not to be viewed as absolute or non-negotiable. However, the core acquisition contract was negotiated with certain storage assumptions which effect costs of storage. Deviations from those assumptions which significantly effect costs would require approval by DOE-Idaho. If some stated requirements are too restrictive, modifications based on sound engineering principles may be negotiated with INEL. 11 figures

  9. CLASSIFICATION OF THE MGR WASTE HANDLING BUILDING VENTILATION SYSTEM

    International Nuclear Information System (INIS)

    J.A. Ziegler

    2000-01-01

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) waste handling building ventilation system structures, systems and components (SSCs) performed by the MGR Preclosure Safety and Systems Engineering Section. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 2000). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 2000). This QA classification incorporates the current MGR design and the results of the ''Design Basis Event Frequency and Dose Calculation for Site Recommendation'' (CRWMS M andO 2000a) and ''Bounding Individual Category 1 Design Basis Event Dose Calculation to Support Quality Assurance Classification'' (Gwyn 2000)

  10. Cask and plug handling system design in port cell

    International Nuclear Information System (INIS)

    Martins, Jean-Pierre; Friconneau, Jean-Pierre; Gabellini, Eros; Keller, Delphine; Levesy, Bruno; Selvi, Anna; Tesini, Alessandro; Utin, Yuri; Wagrez, Julien

    2011-01-01

    The ITER maintenance strategy relies partly on the remote transfer of components from vacuum vessel to hot cells. This function will be fulfilled by transfer cask systems. This paper describes the recent design progresses on interfaces in order to increase components handling feasibility by implementing continuous guiding features that avoid cantilevered loads on the in-cask tractor. Also the design has progressed in order to allow generic docking of the casks. When the cask is connected to the port, it becomes part of the machine first confinement boundary, thus it must provide tightness continuity. This high level safety function was one of the main concerns of a finite element analysis study that has been performed to assess the behavior of the whole system. Numerical analysis methodology and results are explained and shown in order to highlight how it has reinforced the knowledge of the system.

  11. Automated cassette-to-cassette substrate handling system

    Science.gov (United States)

    Kraus, Joseph Arthur; Boyer, Jeremy James; Mack, Joseph; DeChellis, Michael; Koo, Michael

    2014-03-18

    An automated cassette-to-cassette substrate handling system includes a cassette storage module for storing a plurality of substrates in cassettes before and after processing. A substrate carrier storage module stores a plurality of substrate carriers. A substrate carrier loading/unloading module loads substrates from the cassette storage module onto the plurality of substrate carriers and unloads substrates from the plurality of substrate carriers to the cassette storage module. A transport mechanism transports the plurality of substrates between the cassette storage module and the plurality of substrate carriers and transports the plurality of substrate carriers between the substrate carrier loading/unloading module and a processing chamber. A vision system recognizes recesses in the plurality of substrate carriers corresponding to empty substrate positions in the substrate carrier. A processor receives data from the vision system and instructs the transport mechanism to transport substrates to positions on the substrate carrier in response to the received data.

  12. KNOWLEDGE-BASED ROBOT VISION SYSTEM FOR AUTOMATED PART HANDLING

    Directory of Open Access Journals (Sweden)

    J. Wang

    2012-01-01

    Full Text Available

    ENGLISH ABSTRACT: This paper discusses an algorithm incorporating a knowledge-based vision system into an industrial robot system for handling parts intelligently. A continuous fuzzy controller was employed to extract boundary information in a computationally efficient way. The developed algorithm for on-line part recognition using fuzzy logic is shown to be an effective solution to extract the geometric features of objects. The proposed edge vector representation method provides enough geometric information and facilitates the object geometric reconstruction for gripping planning. Furthermore, a part-handling model was created by extracting the grasp features from the geometric features.

    AFRIKAANSE OPSOMMING: Hierdie artikel beskryf ‘n kennis-gebaseerde visiesisteemalgoritme wat in ’n industriёle robotsisteem ingesluit word om sodoende intelligente komponenthantering te bewerkstellig. ’n Kontinue wasige beheerder is gebruik om allerlei objekinligting deur middel van ’n effektiewe berekeningsmetode te bepaal. Die ontwikkelde algoritme vir aan-lyn komponentherkenning maak gebruik van wasige logika en word bewys as ’n effektiewe metode om geometriese inligting van objekte te bepaal. Die voorgestelde grensvektormetode verskaf voldoende inligting en maak geometriese rekonstruksie van die objek moontlik om greepbeplanning te kan doen. Voorts is ’n komponenthanteringsmodel ontwikkel deur die grypkenmerke af te lei uit die geometriese eienskappe.

  13. Optimized hardware design for the divertor remote handling control system

    Energy Technology Data Exchange (ETDEWEB)

    Saarinen, Hannu [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland)], E-mail: hannu.saarinen@tut.fi; Tiitinen, Juha; Aha, Liisa; Muhammad, Ali; Mattila, Jouni; Siuko, Mikko; Vilenius, Matti [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Jaervenpaeae, Jorma [VTT Systems Engineering, Tekniikankatu 1, 33720 Tampere (Finland); Irving, Mike; Damiani, Carlo; Semeraro, Luigi [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2009-06-15

    A key ITER maintenance activity is the exchange of the divertor cassettes. One of the major focuses of the EU Remote Handling (RH) programme has been the study and development of the remote handling equipment necessary for divertor exchange. The current major step in this programme involves the construction of a full scale physical test facility, namely DTP2 (Divertor Test Platform 2), in which to demonstrate and refine the RH equipment designs for ITER using prototypes. The major objective of the DTP2 project is the proof of concept studies of various RH devices, but is also important to define principles for standardizing control hardware and methods around the ITER maintenance equipment. This paper focuses on describing the control system hardware design optimization that is taking place at DTP2. Here there will be two RH movers, namely the Cassette Multifuctional Mover (CMM), Cassette Toroidal Mover (CTM) and assisting water hydraulic force feedback manipulators (WHMAN) located aboard each Mover. The idea here is to use common Real Time Operating Systems (RTOS), measurement and control IO-cards etc. for all maintenance devices and to standardize sensors and control components as much as possible. In this paper, new optimized DTP2 control system hardware design and some initial experimentation with the new DTP2 RH control system platform are presented. The proposed new approach is able to fulfil the functional requirements for both Mover and Manipulator control systems. Since the new control system hardware design has reduced architecture there are a number of benefits compared to the old approach. The simplified hardware solution enables the use of a single software development environment and a single communication protocol. This will result in easier maintainability of the software and hardware, less dependence on trained personnel, easier training of operators and hence reduced the development costs of ITER RH.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. 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

  16. Status report, canister fabrication

    International Nuclear Information System (INIS)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika; Emilsson, Goeran

    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

  17. Research on Handling Stability of Steering-by-wire System

    Directory of Open Access Journals (Sweden)

    Yuan Ying

    2017-01-01

    Full Text Available The main function of steer-by-wire (SBW system are improving steering characteristics, security and stability of the vehicle. In this paper, the variable steering ratio of SBW system is analyzed, and the method of steering ratio based on fuzzy control and neural network are researched. In order to solve the actual working condition, the wheel angle may not reach the expected value, this paper establishes a twodegree-of-freedom (2-DOF vehicle model, and a Matlab/Simulink simulation model, in which a control strategy based on PID controller is put forward to control the front wheel steering angle. Simulation results show that proposed control strategy based on fuzzy neural network can effectively reduce lateral deviation and improve the handling stability and comfortability of the vehicle.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  19. K West Basin canister survey

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1998-01-01

    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

  20. 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].

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  2. Friction welded closures of waste canisters

    International Nuclear Information System (INIS)

    Klein, R.F.

    1987-01-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it into a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive-seal weld, the properties and thickness of which must be at least equal to those of the canister material. All studies and tests performed in the work discussed in this paper have the inertia friction welding concept to be highly feasible in this application. This paper describes the decision to investigate the inertia friction welding process, the inertia friction welding process itself, and a proposed equipment design concept. This system would provide a positive, reliable, inspectable, and full-thickness seal weld while utilizing easily maintainable equipment. This high-quality weld can be achieved even in highly contaminated hot cell

  3. Conceptual design of Blanket Remote Handling System for CFETR

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  4. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  5. 20 CFR 658.401 - Types of complaints handled by the JS complaint system.

    Science.gov (United States)

    2010-04-01

    ... 20 Employees' Benefits 3 2010-04-01 2010-04-01 false Types of complaints handled by the JS... § 658.401 Types of complaints handled by the JS complaint system. (a)(1) The types of complaints (JS related complaints) which shall be handled to resolution by the JS complaint system are as follows: (i...

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

    International Nuclear Information System (INIS)

    Tanaka, K.H.; Minakawa, M.; Yamanoi, Y.

    1991-01-01

    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)

  7. Preliminary design of the high-level waste canister storage system: Topical report for the period of January 1, 1987--September 30, 1987

    International Nuclear Information System (INIS)

    Peters, F.E.; Leap, D.R.

    1987-11-01

    The final stage of the West Valley solidification program will be to place the high-level waste canisters in interim storage until a federal repository is ready to receive them. The waste canisters will be stored in the largest former fuel reprocessing cell at West Valley modified for this purpose. This report provides a description of the preliminary design of the Waste Canister Storage Facility. 9 refs., 14 figs., 1 tab

  8. 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

  9. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

    Stringer, J.

    1992-11-01

    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.)

  10. Corrosion resistance of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban

    2007-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. 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, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  11. Corrosion resistance of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban [Corrosion and Metals Research Institute, Sto ckholm (Sweden)

    2007-03-15

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. 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, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  12. National safeguards system operations at a bulk-handling facility

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The presentation centers on the State System of Accounting and Control (SSAC) for bulk-handling facilities in the licenses sector of the US nuclear community. Details of those material control and accounting measures dealing with the national safeguards program are discussed in Session 6a. The concept and role of the Fundamental Nuclear Material Control (FNMC) Plan are discussed with the participants. In Session 6b, the lecture focusses on the international safeguards program of the US SSAC. The relationship of the national and international requirements is discussed as they relate to the IAEA INFCIRC/153 document. The purpose of this session is to enable participants to: (1) understand the basic MC and A elements in an SSAC; (2) understand which MC and A elements serve the country's national interests and those that serve IAEA safeguards

  13. 324 Building liquid waste handling and removal system project plan

    Energy Technology Data Exchange (ETDEWEB)

    Ham, J.E.

    1998-07-29

    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.

  14. Non-POSIX File System for LHCb Online Event Handling

    CERN Document Server

    Garnier, J C; Cherukuwada, S S

    2011-01-01

    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 a few days for technical maintenance or 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 file-system such as NFS or CIFS we have implemented a custom, light-weight, non-Posix network file-system for the handling of these files. Streaming this file-system 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 fail-over of the read and write service. The writing part of this streaming service is in successful use for the Online, real-time writing of the d...

  15. Maintenance of the JET active gas handling system

    International Nuclear Information System (INIS)

    Brennan, P.D.; Bell, A.C.; Brown, K.; Cole, C.; Cooper, B.; Gibbons, C.; Harris, M.; Jones, G.; Knipe, S.; Lewis, J.; Manning, C.; Miller, A.; Perevezentsev, A.; Skinner, N.; Stagg, R.; Stead, M.; Thomas, R.; Yorkshades, J.

    2003-01-01

    The JET active gas handling system (AGHS) has been in operation in conjunction with the JET machine since Spring 1997. The tritium levels within the vessel have remained sufficiently high, 6.2 g at the end of the DTE1 experiment and currently 1.5 g, such that the AGHS has been required to operate continuously to detritiate gases liberated during D-D operations and to maintain discharges to the environment to ALARP. Maintaining the system to ensure continued operation has been a key factor in guaranteeing the continued availability of the essential sub-systems. The operational history of the JET AGHS has been previously documented in a number of papers [R. Laesser, et al. Proc. of the 19th SOFT Conf. 1 (1996) 227; R. Laesser, et al., Fusion Eng. Des. 46 (1999) 307; P.D. Brennan, et al., 18th Symp. on Fusion Eng., 1999]. Operational downtime is minimised through well-engineered sub-systems that use high integrity components. Outage, contamination and operator dosage are minimised through pre-planned and prepared maintenance operations. The reliability of sub-system critical condition fault detection is demonstrated through routine testing of hard-wired alarms and interlocks

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

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1996-01-01

    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

  17. Development of a telerobotic system for handling contaminated process equipment

    International Nuclear Information System (INIS)

    Fisher, J.J.; Ward, C.R.; Schuler, T.F.

    1987-01-01

    E. I. du Pont de Nemours and Company is evaluating a unique eight-degree-of-freedom Telerobot manipulator to perform size-reduction and material handling operations on contaminated process equipment at the Savannah River Plant (SRP). The Telerobot will be installed in the proposed Transuranic (TRU) Waste Processing Facility, which is scheduled to be operational by 1990. A full-scale prototype Telerobot, manufactured by GCA Corporation, St. Paul, MN is being tested with other process equipment in the Components Test Facility at the Savannah River Laboratory (SRL). All telerobotic operations required in the TRU Waste Facility such as crate unpacking, equipment dismantling, material size-reduction, and selected maintenance operations are being tested. This paper discusses the major mechanical and control features of the Telerobot system. Several system enhancements were added by SRL, including a new quick-hand-change coupling and expanded software control functions. The new software enables a system operator to perform both teleoperated and automatic tasks through several operating modes. These enhancements, as well as future mechanical, control system, and software features, are reviewed

  18. 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.

  19. Underground transportation and handling system for Pollux-casks

    International Nuclear Information System (INIS)

    Schrimpf, C.

    1988-01-01

    The concept for the underground transportation and handling system for Pollux-casks was optimized in a first phase by dividing the process in the repository up into the several transportation and manipulation steps. For each step, the possibilities were described and evaluated by means of a list of criteria (technical, safety and economical criteria). The following concept for the transportation and handling was developed: The casks are transported to the unloading area of the surface facilities by railway or truck. After removal of the transport protection, the entry control is performed. The cask is lifted from the vehicle and placed on a railbound transportation vehicle. This transport unit is transferred to the shaft and placed there ready for shaft hoisting. With the hoisting cage protruding, the transport unit is placed on the hoisting cage by means of a pushing-on device, locked, and then conveyed underground. After arrival on the emplacement level, the transport unit is pulled-off from the hoisting cage and taken over by a mine locomotive and transferred through the transportation and access drifts as far as to the emplacement site. There the locomotive pushed the rail transport vehicle into the emplacement drift, as far as to the designated emplacement position. At the emplacement position, the cask is again lifted by means of hoisting equipment. The rail transport vehicle is pulled out of the emplacement drift and returned to the surface for reloading. After deposition of the cask on the drift floor, the emplacement equipment is pulled back in order to give the operation space free for the slinger backfill truck. Within preceding tests two different backfilling techniques were investigated under realistic conditions: pneumatic backfilling and slinger backfilling. The slinger truck was found to be the most suitable for the designated purpose

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

    International Nuclear Information System (INIS)

    Shen, E.J.; Schweiger, L.J.; Miller, W.C.; Gluck, R.; Devies, S.M.

    1987-04-01

    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

  1. Worklist handling in workflow-enabled radiological application systems

    Science.gov (United States)

    Wendler, Thomas; Meetz, Kirsten; Schmidt, Joachim; von Berg, Jens

    2000-05-01

    For the next generation integrated information systems for health care applications, more emphasis has to be put on systems which, by design, support the reduction of cost, the increase inefficiency and the improvement of the quality of services. A substantial contribution to this will be the modeling. optimization, automation and enactment of processes in health care institutions. One of the perceived key success factors for the system integration of processes will be the application of workflow management, with workflow management systems as key technology components. In this paper we address workflow management in radiology. We focus on an important aspect of workflow management, the generation and handling of worklists, which provide workflow participants automatically with work items that reflect tasks to be performed. The display of worklists and the functions associated with work items are the visible part for the end-users of an information system using a workflow management approach. Appropriate worklist design and implementation will influence user friendliness of a system and will largely influence work efficiency. Technically, in current imaging department information system environments (modality-PACS-RIS installations), a data-driven approach has been taken: Worklist -- if present at all -- are generated from filtered views on application data bases. In a future workflow-based approach, worklists will be generated by autonomous workflow services based on explicit process models and organizational models. This process-oriented approach will provide us with an integral view of entire health care processes or sub- processes. The paper describes the basic mechanisms of this approach and summarizes its benefits.

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

    International Nuclear Information System (INIS)

    Xiong, Rui; Siegel, David; Ross, David

    2014-01-01

    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

  3. 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.

  4. 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

  5. Design, production and initial state of the canister

    International Nuclear Information System (INIS)

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

    2010-12-01

    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

  6. Hazard Control Extensions in a COTS Based Data Handling System

    Science.gov (United States)

    Vogel, Torsten; Rakers, Sven; Gronowski, Matthias; Schneegans, Joachim

    2011-08-01

    EML is an electromagnetic levitator for containerless processing of conductive samples on the International Space Station. This material sciences experiment is running in the European Drawer Rack (EDR) facility. The objective of this experiment is to gain insight into the parameters of liquid metal samples and their crystallisation processes without the influence of container walls. To this end the samples are electromagnetically positioned in a coil system and then heated up beyond their melting point in an ultraclean environment.The EML programme is currently under development by Astrium Space Transportation in Friedrichshafen and Bremen; jointly funded by ESA and DLR (on behalf of BMWi, contract 50WP0808). EML consists of four main modules listed in Table 1. The paper focuses mainly on the architecture and design of the ECM module and its contribution to a safe operation of the experiment. The ECM is a computer system that integrates the power supply to the EML experiment, control functions and video handling and compression features. Experiment control is performed by either telecommand or the execution of predefined experiment scripts.

  7. Development of high intensity beam handling system, 4

    International Nuclear Information System (INIS)

    Yamanoi, Yutaka; Tanaka, Kazuhiro; Minakawa, Michifumi

    1992-01-01

    We have constructed the new counter experimental hall at the KEK 12 GeV Proton Synchrotron (KEK-PS) in order to handle high intensity primary proton beams of up to 1x10 3 pps (protons per second), which is one order of magnitude greater than the present beam intensity of the KEK-PS, 1x10 12 pps. New technologies for handling high-intensity beams have, then, been developed and employed in the construction of the new hall. A part of our R/D work on handling high intensity beams will be reported. (author)

  8. Multi-purpose canister project overview

    International Nuclear Information System (INIS)

    Williams, J.

    1995-01-01

    In this presentation, the author lists the approved and proposed dry storage technologies. He discusses the compatibility of dry storage systems with waste management systems. Historical aspects, recent history, key features of the program approach, benefits, specifications, acquisition and potential utility use of the multi-purpose canister (MPC) are covered. The MPCs provide standardization in the waste management system and a cost savings to utilities and government. MPC will be developed to the same level as existing dry storage systems

  9. Yeast prions: structure, biology, and prion-handling systems.

    Science.gov (United States)

    Wickner, Reed B; Shewmaker, Frank P; Bateman, David A; Edskes, Herman K; Gorkovskiy, Anton; Dayani, Yaron; Bezsonov, Evgeny E

    2015-03-01

    A prion is an infectious protein horizontally transmitting a disease or trait without a required nucleic acid. Yeast and fungal prions are nonchromosomal genes composed of protein, generally an altered form of a protein that catalyzes the same alteration of the protein. Yeast prions are thus transmitted both vertically (as genes composed of protein) and horizontally (as infectious proteins, or prions). Formation of amyloids (linear ordered β-sheet-rich protein aggregates with β-strands perpendicular to the long axis of the filament) underlies most yeast and fungal prions, and a single prion protein can have any of several distinct self-propagating amyloid forms with different biological properties (prion variants). Here we review the mechanism of faithful templating of protein conformation, the biological roles of these prions, and their interactions with cellular chaperones, the Btn2 and Cur1 aggregate-handling systems, and other cellular factors governing prion generation and propagation. Human amyloidoses include the PrP-based prion conditions and many other, more common amyloid-based diseases, several of which show prion-like features. Yeast prions increasingly are serving as models for the understanding and treatment of many mammalian amyloidoses. Patients with different clinical pictures of the same amyloidosis may be the equivalent of yeasts with different prion variants. Copyright © 2015, American Society for Microbiology. All Rights Reserved.

  10. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

    Stringer, J.

    1992-11-01

    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.)

  11. Fuel handling and storage systems in nuclear power plants

    International Nuclear Information System (INIS)

    1984-01-01

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

  12. Storage, transportation and disposal system for used nuclear fuel assemblies

    Science.gov (United States)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  14. Techniques for freeing deposited canisters. Final report

    International Nuclear Information System (INIS)

    Kalbantner, P.; Sjoeblom, R.

    2000-06-01

    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

  15. Evolution of a test article handling system for the SP-100 GES test

    International Nuclear Information System (INIS)

    Shen, E.J.; Schweiger, L.J.; Miller, W.C.; Gluck, R.; Davies, S.M.

    1987-01-01

    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 operation. The test site operator, working in conjunction with 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

  16. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  18. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    International Nuclear Information System (INIS)

    Wensel, R.G.; Metcalfe, R.

    1996-01-01

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

  1. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    Kukkola, T.; Toermaelae, V.P.

    2003-01-01

    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.)

  2. 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.

  3. Criticality safety evaluation report for the Cold Vacuum Drying Facility's process water handling system

    International Nuclear Information System (INIS)

    Roblyer, S.D.

    1998-01-01

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility (CVDF). The controls and limitations on equipment design and operations to control potential criticality occurrences are identified. The effectiveness of equipment design and operation controls in preventing criticality occurrences during normal and abnormal conditions is evaluated and documented in this report. Spent nuclear fuel (SNF) is removed from existing canisters in both the K East and K West Basins and loaded into a multicanister overpack (MCO) in the K Basin pool. The MCO is housed in a shipping cask surrounded by clean water in the annulus between the exterior of the MCO and the interior of the shipping cask. The fuel consists of spent N Reactor and some single pass reactor fuel. The MCO is transported to the CVDF near the K Basins to remove process water from the MCO interior and from the shipping cask annulus. After the bulk water is removed from the MCO, any remaining free liquid is removed by drawing a vacuum on the MCO's interior. After cold vacuum drying is completed, the MCO is filled with an inert cover gas, the lid is replaced on the shipping cask, and the MCO is transported to the Canister Storage Building. The process water removed from the MCO contains fissionable materials from metallic uranium corrosion. The process water from the MCO is first collected in a geometrically safe process water conditioning receiver tank. The process water in the process water conditioning receiver tank is tested, then filtered, demineralized, and collected in the storage tank. The process water is finally removed from the storage tank and transported from the CVDF by truck

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

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1999-01-01

    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

  5. Transport of multiassembly sealed canisters

    International Nuclear Information System (INIS)

    Quinn, R.D.; Lehnert, R.A.; Rosa, J.M.

    1992-01-01

    A significant portion of the commercial spent nuclear fuel in dry storage in the US will be stored in multiassembly sealed canisters before the DOE begins accepting fuel from utilities in 1998. This paper reports that it is desirable from economic and ALARA perspectives to transfer these canisters directly from the plant to the MRS. To this end, it is necessary that the multiassembly sealed canisters, which have been licensed for storage under 10CFR72, be qualified for shipment within a suitable shipping cask under the rules of 10CFR71. Preliminary work performed to date indicates that it is feasible to license a current canister design for transportation, and work is proceeding on obtaining NRC approval

  6. The Sample Handling System for the Mars Icebreaker Life Mission: from Dirt to Data

    Science.gov (United States)

    Dave, Arwen; Thompson, Sarah J.; McKay, Christopher P.; Stoker, Carol R.; Zacny, Kris; Paulsen, Gale; Mellerowicz, Bolek; Glass, Brian J.; Wilson, David; Bonaccorsi, Rosalba; hide

    2013-01-01

    The Mars icebreaker life mission will search for subsurface life on mars. It consists of three payload elements: a drill to retrieve soil samples from approx. 1 meter below the surface, a robotic sample handling system to deliver the sample from the drill to the instruments, and the instruments themselves. This paper will discuss the robotic sample handling system.

  7. Comments on 'SKB RD and D-Programme 98'. Focused on canister integrity and corrosion

    International Nuclear Information System (INIS)

    Bowyer, W.H.; Hermansson, H.P.

    1999-04-01

    level of metallurgical support is required. We disagree that suitable full size canisters have already 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 and handling procedures for the canister system that have to be demonstrated not to be harmful. We feel that there are many areas, which need further evaluation but are granted too little space in the programme. This is valid for i.a. effects of non-uniform loading and creep, welding, quality control, effects of radiolysis and corrosion properties. We also consider that more information should be provided on the detail and timing of the development plan for the trial fabrication programme of the canister, the canister test programme, determination of quality standards and development of non destructive testing procedures. We also feel that insufficient emphasis has been placed on the further development on alternatives to high power electron beam welding, non-destructive testing and over all handling. Copper will be exposed for both general and different kinds of localised corrosion in the repository. The complex mechanical, chemical and microbial environment with high pressures varying in time and location and with oxygen, chloride, sulphur and carbon bearing compounds present will cause different types of attacks that are going to prevail during different time periods. The procedures of production, handling and treatment of the canister throughout the processes of filling, transportation and deposition are crucial for its later, corrosion related integrity throughout the storage period in the repository. There is a risk that due to systematically induced faults, many canisters may have later corrosion related problems. The QA system should be developed to cover all steps of canister handling. We feel a large uncertainty when expressions like 'known corrosion processes

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

    International Nuclear Information System (INIS)

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

    1982-10-01

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

  9. Reliability in sealing of canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Ronneteg, Ulf; Cederqvist, Lars; Ryden, Haakan; Oeberg, Tomas; Mueller, Christina

    2006-06-01

    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

  10. 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.

  11. Dry cask handling system for shipping nuclear fuel

    International Nuclear Information System (INIS)

    Jones, C.R.

    1975-01-01

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

  12. 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)

    Wu Ping; Stepinski, T.

    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

  13. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    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

  14. Contact-handled transuranic transportation system structural analysis

    International Nuclear Information System (INIS)

    Lamoreaux, G.H.; Romesberg, L.E.; Sutherland, S.H.; Duffey, T.A.

    1980-01-01

    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

  15. Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

    International Nuclear Information System (INIS)

    Allen, G.K.

    1980-08-01

    Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

  16. Generic Planning and Control of Automated Material Handling Systems: Practical Requirements Versus Existing Theroy.

    NARCIS (Netherlands)

    Haneyah, S.W.A.; Zijm, Willem H.M.; Schutten, Johannes M.J.; Schuur, Peter

    2011-01-01

    This paper discusses the problem of generic planning and control of Automated Material Handling Systems (AMHSs). The paper illustrates the relevance of this research direction, and then addresses three different market sectors where AMHSs are used. These market sectors are: baggage handling,

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

    NARCIS (Netherlands)

    Haneyah, S.W.A.; Hurink, Johann L.; Schutten, Johannes M.J.; Zijm, Willem H.M.; Schuur, Peter; Hu, Bo; Morasch, Karl; Pickl, Stefan; 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,

  18. Generic planning and control of automated material handling systems : practical requirements versus existing theory

    NARCIS (Netherlands)

    Haneyah, S.W.A.; Schutten, Johannes M.J.; Schuur, Peter; Zijm, Willem H.M.

    2013-01-01

    This paper discusses the problem to design a generic planning and control architecture for utomated material handling systems (AMHSs). We illustrate the relevance of this research direction, and then address three different market sectors where AMHSs are used, i.e., baggage handling, distribution,

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  20. The use of physical model simulation to emulate an AGV material handling system

    International Nuclear Information System (INIS)

    Hurley, R.G.; Coffman, P.E.; Dixon, J.R.; Walacavage, J.G.

    1987-01-01

    This paper describes an application of physical modeling to the simulation of a prototype AGV (Automatic Guided Vehicle) material handling system. Physical modeling is the study of complex automated manufacturing and material handling systems through the use of small scale components controlled by mini and/or microcomputers. By modeling the mechanical operations of the proposed AGV material handling system, it was determined that control algorithms and AGV dispatch rules could be developed and evaluated. This paper presents a brief explanation of physical modeling as a simulation tool and addresses in detail the development of the control algorithm, dispatching rules, and a prototype physical model of a flexible machining system

  1. Development of remote handling system based on 3-D shape recognition technique

    International Nuclear Information System (INIS)

    Tomizuka, Chiaki; Takeuchi, Yutaka

    2006-01-01

    In a nuclear facility, the maintenance and repair activities must be done remotely in a radioactive environment. Fuji Electric Systems Co., Ltd. has developed a remote handling system based on 3-D recognition technique. The system recognizes the pose and position of the target to manipulate, and visualizes the scene with the target in 3-D, enabling an operator to handle it easily. This paper introduces the concept and the key features of this system. (author)

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

    International Nuclear Information System (INIS)

    Treutterer, W.; Neu, G.; Rapson, C.; Raupp, G.; Zasche, D.; Zehetbauer, T.

    2013-01-01

    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

    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

  4. 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)

  5. Handling medical negligence: necessity of a proper system

    Directory of Open Access Journals (Sweden)

    Nuwadatta Subedi

    2017-12-01

    field by the councils framed to settle the cases of medical negligence, especially the consumer protection council, Nepal Medical Council, etc.It is a high time for the government to play role to address to this sensitive and quai-legal issue. If this trend is not ended, the trust of patients towards the doctors may nullify. The doctors also cannot work effectively when there is no safe working environment and may begin to have tendency to abandon handling critical cases with a fear of vulnerability to defamation if not physical assault in worst case. The situation is ultimately disastrous to the entire health care system. The very sensitive issue of health care delivery cannot be compared to any other commercial issues. The proper mechanism of addressing the issue of medical negligence should be practiced alongside encouragement from governmental urge to come up with better medicolegal systems in this regard. This is highly essential for balancing expectations of the patients and performance of medical professionals.

  6. 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.

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

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    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

  8. Development of remote handling techniques for the HLLW solidification plant

    International Nuclear Information System (INIS)

    Tosha, Yoshitsugu; Iwata, Toshio; Inada, Eiichi; Nagaki, Hiroshi; Yamamoto, Masao

    1982-01-01

    To develop the techniques for the remote maintenance of the equipment in a HLLW (high-level liquid waste) solidification plant, the mock-up test facility (MTF) has been designed and constructed. Before its construction, the specific mock-up equipment was manufactured and tested. The results of the test and the outline of the MTF are described. As the mock-up equipment, a denitrater-concentrator, a ceramic melter and a canister handling equipment were selected. Remote operation was performed according to the maintenance program, and the evaluation of the component was conducted on the easiness of operation, performance, and the suitability to remote handling equipment. As a result of the test, four important elements were identified; they were guides, lifting fixtures, remote handling bolts, and remote pipe connectors. Many improvements of these elements were achieved, and reflected in the design of the MTF. The MTF is a steel-framed and slate-covered building (25 mL x 20 mW x 27 mH) with five storys of test bases. It contains the following four main systems: pretreatment and off-gas treatment system, glass melting system, canister handling system and secondary waste liquid recovery system. Further development of the remote maintenance techniques is expected through the test in the MTF. (Aoki, K.)

  9. Simulator for candu600 fuel handling system. environmental implications

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. Improvement Of Physical Ergonomics Using Material Handling Systems

    Directory of Open Access Journals (Sweden)

    Naveen Kumar

    2017-01-01

    Full Text Available This research paper is an investigation of the physical ergonomics of the work place in an automotive parts manufacturing company . Material transfer from one station to another station was done by hand including a walk of a few steps to the next station. The unmachined components that has a quite heavy raw weight also they are being loaded and unloaded by hand .Due to this continuous practice some workers began complaining physical pain in their backs and muscular related pains. The work conditions of the workers were assessed using the REBA Rapid Entire Body Assessment test to understand the stress and the impact the work environment they are exposed to. Few material handling concepts have been suggested and explained to improve the quality of the work conditions for the workers and the REBA test tends to show some significant improvement when these improvements are implemented into the production line.

  11. Combined application of Product Lifecycle and Software Configuration Management systems for ITER remote handling

    International Nuclear Information System (INIS)

    Muhammad, Ali; Esque, Salvador; Aha, Liisa; Mattila, Jouni; Siuko, Mikko; Vilenius, Matti; Jaervenpaeae, Jorma; Irving, Mike; Damiani, Carlo; Semeraro, Luigi

    2009-01-01

    The advantages of Product Lifecycle Management (PLM) systems are widely understood among the industry and hence a PLM system is already in use by International Thermonuclear Experimental Reactor (ITER) Organization (IO). However, with the increasing involvement of software in the development, the role of Software Configuration Management (SCM) systems have become equally important. The SCM systems can be useful to meet the higher demands on Safety Engineering (SE), Quality Assurance (QA), Validation and Verification (V and V) and Requirements Management (RM) of the developed software tools. In an experimental environment, such as ITER, the new remote handling requirements emerge frequently. This means the development of new tools or the modification of existing tools and the development of new remote handling procedures or the modification of existing remote handling procedures. PLM and SCM systems together can be of great advantage in the development and maintenance of such remote handling system. In this paper, we discuss how PLM and SCM systems can be integrated together and play their role during the development and maintenance of ITER remote handling system. We discuss the possibility to investigate such setup at DTP2 (Divertor Test Platform 2), which is the full scale mock-up facility to verify the ITER divertor remote handling and maintenance concepts.

  12. New handling systems as technical support for the working process. Part 6. Feeding devices

    Energy Technology Data Exchange (ETDEWEB)

    Becher, H; Burkhardt, R; Drexel, P; Graf, B; Krreis, W

    1982-03-01

    Social, technical and economic reasons require an enhanced application of handling systems such as industrial robots. Quality and efficiency of an industrial robot depends greatly on feeding devices, and the ARGE-HHS within its project new handling systems as a technical aid in the working process intends to analyze all feeding devices that are likely to be most suitable for advanced applications. Forty one feeding devices were developed, known devices were modified, adapted to different applications, and tested. A variety of feeding devices for most applications in the field of material handling is reported.

  13. Autonomous underwater handling system for service, measurement and cutting tasks for the decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Hahn, M.; Haferkamp, H.; Bach, W.; Rose, N.

    1992-01-01

    For about 10 years the Institute for Material Science at the Hanover University has worked on projects of underwater cutting and welding. Increasing tasks to be done in nuclear facilities led to the development of special handling systems to support and handle the cutting tools. Also sensors and computers for extensive and complex tasks were integrated. A small sized freediving handling system, equipped with 2 video cameras, ultrasonic and radiation sensors and a plasma cutting torch for inspection and decommissioning tasks in nuclear facilities is described in this paper. (Author)

  14. Notification: Audit of Security Categorization for EPA Systems That Handle Hazardous Material Information

    Science.gov (United States)

    Project #OA-FY18-0089, January 8, 2018. The OIG plans to begin preliminary research to determine whether the EPA classified the sensitivity of data for systems that handle hazardous waste material information as prescribed by NIST.

  15. Gas-handling system for studies of tritium-containing materials

    International Nuclear Information System (INIS)

    Carstens, D.H.W.

    1975-01-01

    A gas handling system for preparation and study of tritium containing compounds and materials is described. The system at any one time can handle amounts of DT gas up to about 3 moles and has provisions for purification, storage, and measurement of the gas. Experimental conditions covering the ranges 20 to 800 0 C and 0.1 Pa to 137 MPa (10 -2 torr to 20,000 psi) can be maintained. (auth)

  16. Conceptual design of the handling and storage system for spent target vessel

    Energy Technology Data Exchange (ETDEWEB)

    Adachi, Junichi; Sasaki, Shinobu; Kaminaga, Masanori; Hino, Ryutaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    A conceptual design of a handling and storage system for spent target vessels has been carried out, in order to establish spent target technology for the neutron scattering facility. The spent target vessels must be treated remotely with high reliability and safety, since they are highly activated and contain the poisonous mercury. The system is composed of a target exchange trolley to exchange the target vessel, remote handling equipment such as manipulators, airtight casks for the spent target vessel, storage pits and so on. This report presents the results of conceptual design study on a basic plan, a handling procedure, main devices and their arrangement of a handling and storage system for the spent target vessels. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2001-05-15

    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. Revision 1 documented 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 in section 3

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

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2001-01-01

    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. Revision 1 documented 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 in section 3.1.5 and will be

  19. Reactor helium system, design specification, operation and handling

    International Nuclear Information System (INIS)

    Badrljica, R.

    1984-06-01

    Apart from detailed design specification of the helium cover gas system of the Ra reactor, this document includes description of the operating regime, instructions for manipulations in the system with the aim of achieving and maintaining stationary gas circulation [sr

  20. Defining Handling Qualities of Unmanned Aerial Systems, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Unmanned Air Systems (UAS) are here to stay and operators are demanding access to the National Airspace System (NAS) for a wide variety of missions. This includes a...

  1. Inorganic analyses of volatilized and condensed species within prototypic Defense Waste Processing Facility (DWPF) canistered waste

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1992-01-01

    The high-level radioactive waste currently stored in carbon steel tanks at the Savannah River Site (SRS) will be immobilized in a borosilicate glass in the Defense Waste Processing Facility (DWPF). The canistered waste will be sent to a geologic repository for final disposal. The Waste Acceptance Preliminary Specifications (WAPS) require the identification of any inorganic phases that may be present in the canister that may lead to internal corrosion of the canister or that could potentially adversely affect normal canister handling. During vitrification, volatilization of mixed (Na, K, Cs)Cl, (Na, K, Cs) 2 SO 4 , (Na, K, Cs)BF 4 , (Na, K) 2 B 4 O 7 and (Na,K)CrO 4 species from glass melt condensed in the melter off-gas and in the cyclone separator in the canister pour spout vacuum line. A full-scale DWPF prototypic canister filled during Campaign 10 of the SRS Scale Glass Melter was sectioned and examined. Mixed (NaK)CI, (NaK) 2 SO 4 , (NaK) borates, and a (Na,K) fluoride phase (either NaF or Na 2 BF 4 ) were identified on the interior canister walls, neck, and shoulder above the melt pour surface. Similar deposits were found on the glass melt surface and on glass fracture surfaces. Chromates were not found. Spinel crystals were found associated with the glass pour surface. Reference amounts of the halides and sulfates were found retained in the glass and the glass chemistry, including the distribution of the halides and sulfates, was homogeneous. In all cases where rust was observed, heavy metals (Zn, Ti, Sn) from the cutting blade/fluid were present indicating that the rust was a reaction product of the cutting fluid with glass and heat sensitized canister or with carbon-steel contamination on canister interior. Only minimal water vapor is present so that internal corrosion of the canister, will not occur

  2. A maintenance support system with document handling capability

    International Nuclear Information System (INIS)

    Fukumoto, A.; Tsumura, K.; Fujii, M.; Tai, I.; Makimo, M.; Watanabe, T.

    1990-01-01

    An operation and maintenance support system, called 'Advanced Man-Machine System for Nuclear Power Plants' (MMS-NPP) is under development with the support of the Japanese Government. Taking full advantage of Artificial Intelligence technology, the system aims to enhance the capability of already developed 'Computerized Operator Support System (COSS)' and gives wider and more advanced support for operation and maintenance. With a brief overview of MMS-NPP, this paper describes a support system for plant patrol and equipment inspection. The system gives guidance for plant patrol and for equipment inspection and provides easy access to plant drawings and documents. A unique knowledge acquisition method, utilizing image processing technology, was proposed in building the system

  3. 49 CFR 232.609 - Handling of defective equipment with ECP brake systems.

    Science.gov (United States)

    2010-10-01

    ... (ECP) Braking Systems § 232.609 Handling of defective equipment with ECP brake systems. (a) Ninety-five... systems. 232.609 Section 232.609 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION BRAKE SYSTEM SAFETY STANDARDS FOR FREIGHT...

  4. 40 CFR 65.161 - Continuous records and monitoring system data handling.

    Science.gov (United States)

    2010-07-01

    ... section. (D) Owners and operators shall retain the current description of the monitoring system as long as... Routing to a Fuel Gas System or a Process § 65.161 Continuous records and monitoring system data handling...) Monitoring system breakdowns, repairs, preventive maintenance, calibration checks, and zero (low-level) and...

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

    Energy Technology Data Exchange (ETDEWEB)

    Queral, V., E-mail: vicentemanuel.queral@ciemat.es [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Urbon, J. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Garcia, A.; Cuarental, I.; Mota, F. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Micciche, G. [CR ENEA Brasimone, I-40035 Camugnano (BO) (Italy); Ibarra, A. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Rapisarda, D. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Casal, N. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain)

    2011-10-15

    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.

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

    International Nuclear Information System (INIS)

    Queral, V.; Urbon, J.; Garcia, A.; Cuarental, I.; Mota, F.; Micciche, G.; Ibarra, A.; Rapisarda, D.; Casal, N.

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    PICKETT, W.W.

    2000-09-22

    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.

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

    International Nuclear Information System (INIS)

    PICKETT, W.W.

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

    Hourfar, D.

    1989-01-01

    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.) [de

  10. Mechanical design of the storage tubes in the HWVP canister storage building

    International Nuclear Information System (INIS)

    Divona, C.J.; Fages, R.; Janicek, G.P.; Mullally, J.A.

    1993-01-01

    Canisters of high-level waste from the Hanford Waste Vitrification Plant (HWVP) will be stored in an adjacent facility, the Canister Storage Building (CSB). The canisters are stored vertically in an array of tubes within the shielded vault area of the CSB. This paper describes the mechanical design of the storage tubes, the shield floor plugs that confine the waste within the tubes and the impact absorber system used to assure that the canisters are not breached in the event of an accidental drop. Installation and testing of the components is also discussed

  11. Reconfigurable materials handling system incorporating part tracking, routing and scheduling

    CSIR Research Space (South Africa)

    Naidu, P

    2006-07-01

    Full Text Available . Transmission range of 10m-30m indoors is common. RFID is a commonly available system which uses either low-cost passive Radio tags, or higher cost active tags, that an RF receiver can then read. An RFID system comprises of a reader, its associated antenna...]. 3. Proposed Tracking System The proposed tracking system consists of a two phases. In the first phase a passive radio frequency (RF) tag is read by a RF reader. In the second phase the information obtained from the reader is wirelessly...

  12. Safety assessment of a robotic system handling nuclear material

    International Nuclear Information System (INIS)

    Atcitty, C.B.; Robinson, D.G.

    1996-01-01

    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

  13. 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.

  14. Measurement and control system for the ITER remote handling mock-up test

    International Nuclear Information System (INIS)

    Oka, K.; Kakudate, S.; Takiguchi, Y.; Ako, K.; Taguchi, K.; Tada, E.; Ozaki, F.; Shibanuma, K.

    1998-01-01

    The mock-up test platforms composed of full-scale remote handling (RH) equipment were developed for demonstrating remote replacement of the ITER blanket and divertor. In parallel, the measurement and control system for operating these RH equipment were constructed on the basis of open architecture with object oriented feature, aiming at realization of fully-remoted automatic operation required for ITER. This paper describes the design concept of the measurement and control system for the remote handling equipment of ITER, and outlines the measured performances of the fabricated measurement system for the remote handling mock-up tests, which includes Data Acquisition System (DAS), Visual Monitoring System (VMS) and Virtual Reality System (VRS). (authors)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Yang, X.; Lim, S.I.; Lee, S.J.; Choi, M.S.

    2006-01-01

    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

  17. 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...

  18. Defining Handling Qualities of Unmanned Aerial Systems, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Unmanned Air Systems (UAS) are no longer coming, they are here, and operators from first responders to commercial operators are demanding access to the National...

  19. The beam handling system of the Oslo Cyclotron

    International Nuclear Information System (INIS)

    Messelt, S.

    1985-11-01

    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

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

    International Nuclear Information System (INIS)

    Cullen, S.; Thom, M.

    1993-01-01

    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)

  1. Choices of canisters and elements for the first fuel shipment from K West Basin

    International Nuclear Information System (INIS)

    Makenas, B.J.

    1995-03-01

    Twenty-two canisters (10 prime and 12 backup candidates) in the K West Basin have been identified as containing fuel which, when examined, will satisfy the Data Quality Objectives for the first fuel shipment from this basin. These were chosen as meeting criteria such as containing relatively long fuel elements, locking bar integrity, and the availability of gas/liquid interface level measurements for associated canister gas traps. Two canisters were identified as having reported broken fuel on initial loading. Usage and interpretation of canister cesium concentration measurements have also been established and levels of maximum and minimum acceptable cesium concentration (from a data optimization point of view) for decapping have been determined although other operational cesium limits may also apply. Criteria for picking particular elements, once a canister is opened, are reviewed in this document. A pristine, a slightly damaged, and a badly damaged element are desired. The latter includes elements with end caps removed but does not include elements which have large amounts of swelling or split cladding that might interfere with handling tools. Finally, operational scenarios have been suggested to aid in the selections of canisters and elements in a way that utilizes anticipated canister gas sampling and leads to a correct and quick choice of elements which will supply the desired data

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

    International Nuclear Information System (INIS)

    Hansford, S.M.; Scott, R.

    1992-01-01

    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)

  3. Concept Design of the Payload Handling Manipulator System. [space shuttle orbiters

    Science.gov (United States)

    1975-01-01

    The design, requirements, and interface definition of a remote manipulator system developed to handle orbiter payloads are presented. End effector design, control system concepts, and man-machine engineering are considered along with crew station requirements and closed circuit television system performance requirements.

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

    International Nuclear Information System (INIS)

    Vinod Kumar, A.; Oza, R.B.; Chaudhury, P.; Suri, M.; Saindane, S.; Singh, K.D.; Bhargava, P.; Sharma, V.K.

    2009-01-01

    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)

  5. Evaluation of a molybdenum assay canister

    International Nuclear Information System (INIS)

    Yoshizumi, T.T.; Keener, S.J.

    1988-01-01

    The performance characteristics of a commercial molybdenum assay canister were evaluated. The geometrical variation of the technetium-99m (/sup 99m/Tc) activity reading was studied as a function of the elution volume for the standard vials. It was found that the /sup 99m/Tc canister activity reading was ∼ 5% lower than that of the standard method. This is due to attenuation by the canister wall. However, the effect of the geometric variation on the clinical dose preparation was found to be insignificant. The molybdenum-99 ( 99 Mo) contamination level was compared by two methods: (1) the commercial canister and (2) the standard assay kit. The 99 Mo contamination measurements with the canister indicated consistently lower readings than those with the standard 99 Mo assay kit. The authors conclude that the canister may be used in the clinical settings. However, the user must be aware of the problems and the limitations associated with this canister

  6. A Gas Target with a Tritium Gas Handling System

    Energy Technology Data Exchange (ETDEWEB)

    Holmqvist, B; Wiedling, T

    1963-12-15

    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.

  7. Handling system for nuclear reactor fuel and reflector elements

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  8. Physicochemical and Geotechnical Alterations to MX-80 Bentonite at the Waste Canister Interface in an Engineered Barrier System

    Directory of Open Access Journals (Sweden)

    Christopher W. Davies

    2017-08-01

    Full Text Available The study investigated the basic geomechanical and mineralogical evolution of the bentonite barrier under various experimental boundary conditions which replicated the near-field Thermo-Hydro-Chemico (THC conditions in a repository. The relationships between the physicochemical alterations and changes in the geotechnical properties have seldom been studied, especially on a consistent dataset. This paper attempts to link the physicochemical properties of Na-bentonite (MX-80 to the macro-scale engineering functionality of the bentonite post THC exposure. Experiments investigated the impact of THC variables on the engineering and physicochemical functionality of the bentonite with respect to its application within a High-Level Waste (HLW engineered barrier system. Intrinsic alterations to the MX-80 bentonite under relatively short-term exposure to hydrothermal and chemical conditions were measured. Additionally, two long-term tests were conducted under ambient conditions to consider the impact of exposure duration. The intrinsic measurements were then related to the overall performance of the bentonite as a candidate barrier material for application in a UK geological disposal facility. Findings indicate that exposure to thermo-saline-corrosion conditions (i.e., corrosion products derived from structural grade 275 carbon steel inhibits the free swell capacity and plasticity of the bentonite. However, the measured values remained above the design limits set out for the Swedish multi-barrier concept, from which the UK concept may take a lead. Corrosion alone does not appear to significantly affect the geotechnical measurements compared with the influence of thermal loading and high saline pore water after relatively short-term exposure. Thermal and corrosion exposure displayed no impact on the intrinsic swelling of the smectite component, indicating that no significant structural alteration had occurred. However, when exploring more complex saline

  9. Remote handling concept for the neutral beam system

    International Nuclear Information System (INIS)

    Choi, Chang-Hwan; Palmer, Jim; Conesa, Carles; Friconneau, Jean-Pierre; Martins, Jean-Pierre; Subramanian, Rajendran; Jeannoutot, Thomas; Graceffa, Joseph; Schunke, Beatrix; Uffelen, MarcoVan; Damiani, Carlo; Tesini, Alessandro

    2011-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Patel, R.; Punshon, C.; Nicholas, J.; Bastid, P.; Zhou, R.; Schneider, C.; Bagshaw, N.; Howse, D.; Hutchinson, E. [TWI Ltd, Cambridge, (United Kingdom); Asano, R. [Hitachi Zosen Corporation, Osaka (Japan); King, S. [Integrity Corrosion Consulting Ltd, Calgary, Alberta (Canada)

    2012-10-15

    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

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

    International Nuclear Information System (INIS)

    Patel, R.; Punshon, C.; Nicholas, J.; Bastid, P.; Zhou, R.; Schneider, C.; Bagshaw, N.; Howse, D.; Hutchinson, E.; Asano, R.; King, S.

    2012-10-01

    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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  13. Operating experiences in fuel handling system at KGS

    International Nuclear Information System (INIS)

    Reddy, G.P.; Nagabhushanam

    2006-01-01

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

  14. Burst Test Qualification Analysis of DWPF Canister-Plug Weld

    International Nuclear Information System (INIS)

    Gupta, N.K.; Gong, Chung.

    1995-02-01

    The DWPF canister closure system uses resistance welding for sealing the canister nozzle and plug to ensure leak tightness. The welding group at SRTC is using the burst test to qualify this seal weld in lieu of the shear test in ASME B ampersand PV Code, Section IX, paragraph QW-196. The burst test is considered simpler and more appropriate than the shear test for this application. Although the geometry, loading and boundary conditions are quite different in the two tests, structural analyses show similarity in the failure mode of the shear test in paragraph QW-196 and the burst test on the DWPF canister nozzle Non-linear structural analyses are performed using finite element techniques to study the failure mode of the two tests. Actual test geometry and realistic stress strain data for the 304L stainless steel and the weld material are used in the analyses. The finite element models are loaded until failure strains are reached. The failure modes in both tests are shear at the failure points. Based on these observations, it is concluded that the use of a burst test in lieu of the shear test for qualifying the canister-plug weld is acceptable. The burst test analysis for the canister-plug also yields the burst pressures which compare favorably with the actual pressure found during burst tests. Thus, the analysis also provides an estimate of the safety margins in the design of these vessels

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

  16. 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.

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

    International Nuclear Information System (INIS)

    Ferreira, João; Vale, Alberto; Ribeiro, Isabel

    2013-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  1. Comments on 'SKB RD and D-Programme 98'. Focused on canister integrity and corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, W.H. [Meadow End Farm, Farnham (United Kingdom); Hermansson, H.P. [Studsvik Material AB, Nykoeping (Sweden)

    1999-04-01

    SKB and that a higher level of metallurgical support is required. We disagree that suitable full size canisters have already 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 and handling procedures for the canister system that have to be demonstrated not to be harmful. We feel that there are many areas, which need further evaluation but are granted too little space in the programme. This is valid for i.a. effects of non-uniform loading and creep, welding, quality control, effects of radiolysis and corrosion properties. We also consider that more information should be provided on the detail and timing of the development plan for the trial fabrication programme of the canister, the canister test programme, determination of quality standards and development of non destructive testing procedures. We also feel that insufficient emphasis has been placed on the further development on alternatives to high power electron beam welding, non-destructive testing and over all handling. Copper will be exposed for both general and different kinds of localised corrosion in the repository. The complex mechanical, chemical and microbial environment with high pressures varying in time and location and with oxygen, chloride, sulphur and carbon bearing compounds present will cause different types of attacks that are going to prevail during different time periods. The procedures of production, handling and treatment of the canister throughout the processes of filling, transportation and deposition are crucial for its later, corrosion related integrity throughout the storage period in the repository. There is a risk that due to systematically induced faults, many canisters may have later corrosion related problems. The QA system should be developed to cover all steps of canister handling. We feel a large uncertainty when expressions like &apos

  2. Model-based design of supervisory controllers for baggage handling systems

    NARCIS (Netherlands)

    Swartjes, L.; van Beek, D.A.; Fokkink, W.J.; van Eekelen, J.A.W.M.

    2017-01-01

    The complexity of airport baggage handling systems in combination with the required high level of robustness makes designing supervisory controllers for these systems a challenging task. We show how a state of the art, formal, model-based design framework has been successfully used for model-based

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

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

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

    International Nuclear Information System (INIS)

    1996-01-01

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

  5. 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...... expansion and two allocation procedures for estimating greenhouse gas (GHG) emissions of molasses ethanol. As seen from our results, system expansion yields the highest estimate among the three. However, no matter which procedure is used, a significant reduction of emissions from the fuel stage...... in the 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...

  6. Development of cold sprayed Cu coating for canister

    International Nuclear Information System (INIS)

    Kim, Hyung Jun; Kang, Yoon Ha

    2010-01-01

    Cold sprayed Cu deposition was studied for the application of outer part of canister for high level nuclear waste. Five commercially available pure Cu powders were analyzed and sprayed by high pressure cold spray system. Electrochemical corrosion test using potentiostat in 3.5% NaCl solution was conducted as well as microstructural analysis including hardness and oxygen content measurements. Overall evaluation of corrosion performance of cold sprayed Cu deposition is inferior to forged and extruded Cu plates, but some of Cu depositions are comparable to Cu plates. The simulated corrosion test in 200m underground cave is still in progress. The effect of cold spray process parameters was also studied and the results show that the type of nozzle is the most important other than powder feed rate, spray distance, and scan speed. 1/10 scale miniature of canister was manufactured confirming that the production of full scale canister is possible

  7. Prototype spent-fuel canister design, analysis, and test

    International Nuclear Information System (INIS)

    Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

    1982-03-01

    Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

  8. 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.

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

    International Nuclear Information System (INIS)

    Sandha, R.S.; Nageswar Rao, J; Dwivedi, Jishnu; Petwal, V.C.; Soni, H.C.

    2005-01-01

    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)

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

    International Nuclear Information System (INIS)

    Herndon, J.N.; Kring, C.T.; Rowe, J.C.

    1986-01-01

    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

  11. Manipulator and materials handling systems for reactor decommissioning -Cooperation between the university and the plant operator

    International Nuclear Information System (INIS)

    Schreck, G.; Bach, F. W.; Haferkamp, H.

    1995-01-01

    Nuclear reactor dismantling requires suitable handling systems for tools and disassembled components, as well as qualified and reliable disassembly and cutting techniques. From the angle of radiation protection, remote-controlled handling techniques and underwater techniques are the methods of choice, the latter particularly in continuation of plant operating conditions, and this all the more the more disassembly work proceeds towards the reactor core. With the experience accumulated for 20 years now by the Institut fuer Werkstoffkunde (materials science) of Hannover University by basic research and application-oriented development work in the field of thermal cutting technology, especially plasma arc cutting techniques, as well as development work in the field of remote-controlled materials handling systems, the institute is the cut-out partner for disassembly tasks in reactor decommissioning. (Orig./DG) [de

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

    Energy Technology Data Exchange (ETDEWEB)

    Raymond, Rick E. [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Frederickson, James R. [AREVA, Avignon (France); Criddle, James [AREVA, Avignon (France); Hamilton, Dennis [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Johnson, Mike W. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

    2012-10-18

    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).

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

    International Nuclear Information System (INIS)

    Raymond, Rick E.; Frederickson, James R.; Criddle, James; Hamilton, Dennis; Johnson, Mike W.

    2012-01-01

    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)

  14. Status of Closure Welding Technology of Canister for Transportation and Storage of High Level Radioactive Material and Waste

    International Nuclear Information System (INIS)

    Lee, H. J.; Bang, K. S.; Seo, K. S.; Seo, C. S.

    2010-10-01

    Closure seal welding is one of the key technologies in fabricating and handling the canister which is used for transportation and storage of high radioactive material and waste. Simple industrial fabrication processes are used before filling the radioactive waste into the canister. But, automatic and remote processes should be used after filling the radioactive material because the thickness of canister is not sufficient to shield the high radiation from filled material or waste. In order to simplify the welding process the closure structure of canister and the sealing method are investigated and developed properly. Two types of radioactive materials such as vitrified waste and compacted solid waste are produced in nuclear industry. Because the filling method of two types of waste is different, the shapes of closure and opening of canister and welding method is also different. The canister shape and sealing method should be standardized to standardize the handling facilities and inspection process such as leak test after closure welding. In order to improve the productivity of disposal and compatibility of the canister, the structure and shape of canister should be standardized considering the type of waste. Two kind of welding process such as arc welding and resistance welding are reported and used in the field. In the arc welding process GTAW and PAW are considered proper processes for closure welding. The closure seal welding process can be selected by considering material of canister, thickness of body, productivity, and applicable codes and rules. Because the storage time of nuclear waste in canister is very long, at least 20 years, the long-time corrosion at the weld should be estimated including mechanical integrity. Recently, the mitigation of residual stress around weld region, which causes stress corrosion cracking, is also interesting research issue

  15. Multi-Canister Overpack (MCO) Topical Report

    International Nuclear Information System (INIS)

    LORENZ, B.D.

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  17. Design premises for canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Werme, L.

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel

  18. Development of Copper Canister through Cold Sprayed Coating Method

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun

    2007-12-15

    General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future.

  19. Development of Copper Canister through Cold Sprayed Coating Method

    International Nuclear Information System (INIS)

    Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun

    2007-12-01

    General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future

  20. Handle system integration as an enabler in an internet of things smart environment

    CSIR Research Space (South Africa)

    Coetzee, L

    2012-06-01

    Full Text Available . The existing main components of the demonstrator are: ? BeachComber: a bearer-agnostic event-driven platform able to receive and send messages via numerous channels (Butgereit & Coetzee, 2011). ? ThingMemory: an enterprise-scale application where a cyber...-hardware platform linked to a number of different sensors and actuators. ? HandleProxy: a newly-created building block acting as proxy to the Handle System. ? Runtime configurable decision engine (embedded within ThingMemory) that, based on received information...

  1. Coal handling system structural analysis for modifications or plant life extension

    International Nuclear Information System (INIS)

    Dufault, A.; Weider, F.; Doyle, P.

    1989-01-01

    One neglected aspect of plant modification or life extension is the extent to which previous projects may have affected the integrity of existing structures. During the course of a project to backfit fire protection facilities to existing coal handling systems, it was found that past modifications had added loads to existing coal handling structures which exceeded the available design margin. This paper describes the studies that discovered the original problem areas, as well as the detailed analysis and design considerations used to repair these structures

  2. Operation of data acquisition and handling system in the INS-SF cyclotron

    International Nuclear Information System (INIS)

    Yasue, M.; Omata, K.

    1976-01-01

    Operations of following data processing routines are described. 1) One-dimensional multiplexer PHA. 2) Two-dimensional multiplexer PHA. 3) Two or three parameter data handling: Digital gating, dumping of raw data onto MT and processing in function modes. These processing routines are executed under the control of a real time disk operating system in TOSBAC-40C. (auth.)

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

    NARCIS (Netherlands)

    van Sinderen, Marten J.; 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

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

    International Nuclear Information System (INIS)

    Ahmed, H.; Cowie, A.; Malek, R.A.; Rafer, A.; Ma, D.; Tebo, F.

    1978-10-01

    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

  5. Application of advanced remote systems technology to future waste handling facilities

    International Nuclear Information System (INIS)

    Kring, C.T.; Meacham, S.A.

    1987-01-01

    The Consolidated Fuel Reprocessing Program (CFRP) at Oak Ridge National Laboratory (ORNL) has been advancing the technology of remote handling and remote maintenance for in-cell systems planned for future nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor is directly applicable to the proposed in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The application of teleoperated, force-reflecting servomanipulators with television viewing could be a major step forward in waste handling facility design. Primary emphasis in the current program is the operation of a prototype remote handling and maintenance system, the advanced servomanipulator (ASM), which specifically addresses the requirements of fuel reprocessing and waste handling with emphasis on force reflection, remote maintainability, reliability, radiation tolerance, and corrosion resistance. Concurrent with the evolution of dexterous manipulators, concepts have also been developed that provide guidance for standardization of the design of the remotely operated and maintained equipment, the interface between the maintenance tools and the equipment, and the interface between the in-cell components and the facility

  6. A simulation-based approach for evaluating logging residue handling systems.

    Science.gov (United States)

    B. Bruce Bare; Benjamin A. Jayne; Brian F. Anholt

    1976-01-01

    Describes a computer simulation model for evaluating logging residue handling systems. The flow of resources is traced through a prespecified combination of operations including yarding, chipping, sorting, loading, transporting, and unloading. The model was used to evaluate the feasibility of converting logging residues to chips that could be used, for example, to...

  7. Multi-Canister overpack sealing configuration

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The Spent Nuclear Fuel (SNF) position regarding the Multi-Canister Overpack (MCO) sealing configuration is to initially rely on an American Society of Mechanical Engineers (ASME) Section III Subsection NB code compliant mechanical closure/sealing system to quickly and safely establish and maintain full confinement of radioactive materials prior to and during MCO fuel drying activities. Previous studies have shown the mechanical seal to be the preferred closure method, based on dose, cost, and schedule considerations. The cost and schedule impacts of redesigning the mechanical closure to a welded shield plug do not support changing the closure system. The SNF Project has determined that the combined mechanical/welded closure system meets or exceeds the regulatory requirements to provide redundant seals while accommodating key safety and schedule limitations that are unique to K Basins fuel removal effort

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

    International Nuclear Information System (INIS)

    Ballard, A.S.

    1977-02-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Adrian, E-mail: adrian.williams@oxfordtechnologies.co.uk [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, Oxon, OX14 1RJ (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Sanders, Stephen [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, Oxon, OX14 1RJ (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Weder, Gerard [Tree-C Technology BV, Buys Ballotstraat 8, 6716 BL Ede (Netherlands); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Bastow, Roger; Allan, Peter; Hazel, Stuart [CCFE, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

    2011-10-15

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

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

    International Nuclear Information System (INIS)

    Williams, Adrian; Sanders, Stephen; Weder, Gerard; Bastow, Roger; Allan, Peter; Hazel, Stuart

    2011-01-01

    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.

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

    International Nuclear Information System (INIS)

    T. A. Misiak

    1998-01-01

    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

  12. A versatile system for the rapid collection, handling and graphics analysis of multidimensional data

    International Nuclear Information System (INIS)

    O'Brien, P.M.; Moloney, G.; O'Oconnor, A.; Legge, G.J.F.

    1991-01-01

    The paper discusses the performances of a versatile computerized system developed at the Microanalytical Research Centre of the Melbourne University, for handling multiparameter data that may arise from a variety of experiments - nuclear, accelerator mass spectrometry, microprobe elemental analysis or 3-D microtomography. Some of the most demanding requirements arise in the application of microprobes to quantitative elemental mapping and to microtomography. A system to handle data from such experiments had been under continuous development. It has been reprogramed to run on a DG DS7540 workstation. The whole system of software has been rewritten, greatly expanded and made much more powerful and faster, by use of modern computer technology - a VME bus computer with a real-time operating system and a RISC workstation running UNIX and the X-window environment

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

    International Nuclear Information System (INIS)

    1977-03-01

    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)

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

    International Nuclear Information System (INIS)

    Dumbreck, A.A.; Murphy, S.P.; Smith, C.W.

    1990-01-01

    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)

  15. 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

    a system based solely on non-economic incentives. The main objective was to compare citizen`s beliefs and attitudes towards waste handling systems and their consequence for motivations to co-operate. Th groups of hypotheses concerning the beliefs-attitude relationship, differences in attitudes between...... 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....

  16. Inspection of disposal canisters components

    International Nuclear Information System (INIS)

    Pitkaenen, J.

    2013-12-01

    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

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

    Directory of Open Access Journals (Sweden)

    Hyo Jik Lee

    2013-10-01

    Full Text Available 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 demonstration facility (PRIDE. Accurate and reliable servo-control is the basic requirement for the BDSM to accomplish any given tasks successfully in a hotcell environment. To achieve this end, the hardware and software of a digital signal processor-based remote control system were fully custom-developed and implemented to control the BDSM. To reduce the residual vibration of the BDSM, several input profiles, including input shaping, were carefully chosen and evaluated. Furthermore, a time delay controller was employed to achieve good tracking performance and systematic gain tuning. The experimental results demonstrate that the applied control algorithms are more effective than conventional approaches. The BDSM successfully completed its performance tests at a mock-up and was installed at PRIDE for real-world operation. The remote handling system at KAERI is expected to advance the actualization of pyroprocessing.

  18. Canister arrangement for storing radioactive waste

    Science.gov (United States)

    Lorenzo, D.K.; Van Cleve, J.E. Jr.

    1980-04-23

    The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

  19. Waste canister for storage of nuclear wastes

    Science.gov (United States)

    Duffy, James B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall.

  20. Waste canister for storage of nuclear wastes

    International Nuclear Information System (INIS)

    Duffy, J.B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall. 4 claims, 4 figures

  1. Progress in the design of the ITER Neutral Beam cell Remote Handling System

    Energy Technology Data Exchange (ETDEWEB)

    Shuff, R., E-mail: robin.shuff@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Van Uffelen, M.; Damiani, C. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Tesini, A.; Choi, C.-H. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Meek, R. [Oxford Technologies Limited, 7 Nuffield Way, Abingdon OX14 1RL (United Kingdom)

    2014-10-15

    The ITER Neutral Beam cell will include a suite of Remote Handling equipment for maintenance tasks. This paper summarises the current status and recent developments in the design of the ITER Neutral Beam Remote Handling System. Its concept design was successfully completed in July 2012 by CCFE in the frame of a grant agreement with F4E, in collaboration with the ITER Organisation, including major systems like monorail crane, Beam Line Transporter, beam source equipment, upper port and neutron shield equipment and associated tooling. Research and development activities are now underway on the monorail crane radiation hardened on-board control system and first of a kind remote pipe and lip seal maintenance tooling for the beam line vessel, reported in this paper.

  2. Track-mounted remote handling system for the Tokamak Fusion Engineering Test

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Daubert, R.L.; Yount, J.A.

    1982-01-01

    Concepts for remote handling machines (IVM) designed to transverse the interior of toroidal vessels with guidance and support from track systems have been developed for the proposed Tokamak Fusion Engineering Test (TFET). TFET has been proposed as an upgrade for the Tokamak Fusion Test Reactor (TFTR), currently nearing completion. The track-mounted IVMs were conceived to perform in-vessel remote maintenance for TFET, including removal and replacement of pump limiter blades and protective tiles as well as other maintenance-related tasks such as vessel wall inspection leak testing and interior cleanup. The conceptual IVMs consist of three manipulator arms mounted on a common frame member: a single power manipulator arm with high load carrying capacity and two lower-capacity servomanipulator arms. Descriptions of the IVM concepts, in-vessel track systems, and ex-vessel handling systems are presented

  3. Progress in the design of the ITER Neutral Beam cell Remote Handling System

    International Nuclear Information System (INIS)

    Shuff, R.; Van Uffelen, M.; Damiani, C.; Tesini, A.; Choi, C.-H.; Meek, R.

    2014-01-01

    The ITER Neutral Beam cell will include a suite of Remote Handling equipment for maintenance tasks. This paper summarises the current status and recent developments in the design of the ITER Neutral Beam Remote Handling System. Its concept design was successfully completed in July 2012 by CCFE in the frame of a grant agreement with F4E, in collaboration with the ITER Organisation, including major systems like monorail crane, Beam Line Transporter, beam source equipment, upper port and neutron shield equipment and associated tooling. Research and development activities are now underway on the monorail crane radiation hardened on-board control system and first of a kind remote pipe and lip seal maintenance tooling for the beam line vessel, reported in this paper

  4. Evaluation of Four Bedside Test Systems for Card Performance, Handling and Safety.

    Science.gov (United States)

    Giebel, Felix; Picker, Susanne M; Gathof, Birgit S

    2008-01-01

    SUMMARY: OBJECTIVE: Pretransfusion ABO compatibility testing is a simple and required precaution against ABO-incompatible transfusion, which is one of the greatest threats in transfusion medicine. While distinct agglutination is most important for correct test interpretation, protection against infectious diseases and ease of handling are crucial for accurate test performance. Therefore, the aim of this study was to evaluate differences in test card design, handling, and user safety. DESIGN: Four different bedside test cards with pre-applied antibodies were evaluated by 100 medical students using packed red blood cells of different ABO blood groups. Criteria of evaluation were: agglutination, labelling, handling, and safety regarding possible user injuries. Criteria were rated subjectively according to German school notes ranging from 1 = very good to 6 = very bad/insufficient. RESULTS: Overall, all cards received very good/good marks. The ABO blood group was identified correctly in all cases. Three cards (no. 1, no. 3, no. 4) received statistically significant (p labelling (1.5 vs. 2.2-2.4), handling (1.9-2.0 vs. 2.5), and user safety (2.5 vs. 3.4). Analysis of card self-explanation revealed no remarkable differences. CONCLUSION: Despite good performance of all card systems tested, the best results when including all criteria evaluated were obtained with card no. 4 (particularly concerning clear agglutination), followed by cards no. 2, no. 1, and no. 3.

  5. Multi-Canister overpack internal HEPA filters

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    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

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

    International Nuclear Information System (INIS)

    Pinna, T.; Caporali, R.; Tesini, A.

    2008-01-01

    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

  7. Advanced dust monitoring system applied to new TRU handling facility of JAERI

    International Nuclear Information System (INIS)

    Yabuta, H.; Shigeta, Y.; Sawahata, K.; Hasegawa, K.

    1993-01-01

    In JAERI, a large, scale multipurpose facility is under construction, which consists of a TRU waste management testing installation, a solution fuel treatment installation and critical assemblies with uranium and/or plutonium solution fuel. The facility is also equipped with a lot of gloveboxes for handling and treatment of solution fuel and hot cells for research on reprocessing process. As there may be a relatively high potential of air contamination, it is important to monitor air contamination effectively and efficiently. An advanced dust monitoring system was introduced for convenience of handling and automatical measurement of filter papers, by developing a filter-holder with an IC memory and a radioactivity measuring device with an automatic filter-holder changing mechanism as a part of a centralized monitoring system with a computer

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

    International Nuclear Information System (INIS)

    Schneider, K.J.; Andrews, W.B.; Schreiber, A.M.; Rosenthal, L.J.; Odle, C.J.

    1981-09-01

    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. The development of a Martian atmospheric Sample collection canister

    Science.gov (United States)

    Kulczycki, E.; Galey, C.; Kennedy, B.; Budney, C.; Bame, D.; Van Schilfgaarde, R.; Aisen, N.; Townsend, J.; Younse, P.; Piacentine, J.

    The collection of an atmospheric sample from Mars would provide significant insight to the understanding of the elemental composition and sub-surface out-gassing rates of noble gases. A team of engineers at the Jet Propulsion Laboratory (JPL), California Institute of Technology have developed an atmospheric sample collection canister for Martian application. The engineering strategy has two basic elements: first, to collect two separately sealed 50 cubic centimeter unpressurized atmospheric samples with minimal sensing and actuation in a self contained pressure vessel; and second, to package this atmospheric sample canister in such a way that it can be easily integrated into the orbiting sample capsule for collection and return to Earth. Sample collection and integrity are demonstrated by emulating the atmospheric collection portion of the Mars Sample Return mission on a compressed timeline. The test results achieved by varying the pressure inside of a thermal vacuum chamber while opening and closing the valve on the sample canister at Mars ambient pressure. A commercial off-the-shelf medical grade micro-valve is utilized in the first iteration of this design to enable rapid testing of the system. The valve has been independently leak tested at JPL to quantify and separate the leak rates associated with the canister. The results are factored in to an overall system design that quantifies mass, power, and sensing requirements for a Martian atmospheric Sample Collection (MASC) canister as outlined in the Mars Sample Return mission profile. Qualitative results include the selection of materials to minimize sample contamination, preliminary science requirements, priorities in sample composition, flight valve selection criteria, a storyboard from sample collection to loading in the orbiting sample capsule, and contributions to maintaining “ Earth” clean exterior surfaces on the orbiting sample capsule.

  11. Graphical models for simulation and control of robotic systems for waste handling

    International Nuclear Information System (INIS)

    Drotning, W.D.; Bennett, P.C.

    1992-01-01

    This paper discusses detailed geometric models which have been used within a graphical simulation environment to study transportation cask facility design and to perform design and analyses of robotic systems for handling of nuclear waste. The models form the basis for a robot control environment which provides safety, flexibility, and reliability for operations which span the spectrum from autonomous control to tasks requiring direct human intervention

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

    International Nuclear Information System (INIS)

    Berg, G.E.

    1977-01-01

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

  13. Treatment of plutonium-contaminated solid waste: a review of handling systems

    International Nuclear Information System (INIS)

    Meredith, B.E.; Hardy, A.R.

    1985-02-01

    Handling techniques are reviewed to identify those suitable for adaptation for use in transporting large items of redundant plutonium contaminated plant and equipment to a remotely operated size reduction facility, moving them into the facility, presenting them to size reduction equipment and loading the processed waste into drums. It is concluded that an integrated system based on a combination of slatted conveyors, roller tables, air transporters and manipulators, merits further consideration. An appropriate experimental programme is outlined. (author)

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

    International Nuclear Information System (INIS)

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

    1983-05-01

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

  15. 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.

  16. Finite element modelling of an evacuated canister for removal of molten radioactive glass

    International Nuclear Information System (INIS)

    Hatchell, B.K.; Deibler, J.E.; Ketner, G.L.

    1994-05-01

    Pacific Northwest Laboratory (PNL) has prepared a preliminary design for the West Valley Demonstration Project evacuated canister system. The function of the evacuated canister is to remove radioactive molten glass from a hot cell melter cavity during a planned melter shutdown. The proposed evacuated canister system consists of an L-shaped 4-inch 304L stainless steel (SS) schedule 40 pipe, sealed at one end with an aluminum plug and attached at the other end to a canister. While the canister is being filled, it is positioned and held above the melter at approximately 15 degree from horizontal by two turntable-mounted cranes. ANSYS finite element analyses were conducted to evaluate the heat transfer from the glass to the canister and establish a maximum canister temperature for material strength evaluation. Finite element structural analyses were conducted to identify areas that required reinforcement for high temperature use. Finite element results will be used to locate strain gauges at high stress locations during prototype testing

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

    International Nuclear Information System (INIS)

    Drotning, W.; Kimberly, H.; Wapman, W.

    1997-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Locke, Darren, E-mail: darren.locke@f4e.europa.eu [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); González Gutiérrez, Carmen; Damiani, Carlo [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Friconneau, Jean-Pierre; Martins, Jean-Pierre [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    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.

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

    International Nuclear Information System (INIS)

    Locke, Darren; González Gutiérrez, Carmen; Damiani, Carlo; Friconneau, Jean-Pierre; Martins, Jean-Pierre

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Adachi, J.; Kakudate, S.; Oka, K.; Seki, M.

    1995-01-01

    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.)

  1. 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.

  2. Study of ITER equatorial port plug handling system and vacuum sealing interface

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Jean-Pierre [Association Euratom CEA, CEA/DSM/IRFM, Cadarache, F-13108 Saint-Paul-lez-Durance (France)], E-mail: jean-pierre.martins@cea.fr; Doceul, Louis; Marol, Sebastien; Delchie, Elise [Association Euratom CEA, CEA/DSM/IRFM, Cadarache, F-13108 Saint-Paul-lez-Durance (France); Cordier, Jean-Jacques; Levesy, Bruno; Tesini, Alessandro [ITER International Organization, F-13108 Saint-Paul-lez-Durance cedex (France); Ciattaglia, Emanuela [EFDA CSU - Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Tivey, Richard [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Gillier, Rene; Abbes, Christophe [GARLOCK - Sealing Technologies - 90, rue de la roche du Geai, F-42029 St-Etienne cedex 1 (France)

    2009-06-15

    In the field of the ITER port plug engineering and integration task, CEA has contributed to define proposals concerning the port plugs vacuum sealing interface with the vessel flange and the equatorial plug handling. The 2001 baseline vacuum flange sealing consisted of TIG welding of a 316L strip plate on to U shapes. This arrangement presented some issues like welding access, implementation of tools, lip consumption, complex local leak test, continuous leak checking. Therefore, an alternate sealing solution based on the use of metallic gaskets is proposed. The different technical aspects are discussed to explain how this design can simplify the maintenance and deal with safety and vacuum requirements. The design of the mechanical attachment and vacuum sealing of the plug has constantly evolved, but the associated remote handling equipment was not systematically reviewed. An update of the cask and maintenance procedure was studied in order to design it in accordance with the last generic plug flange design. This includes a concept of a gripping system that uses the plug flange bolting area and, to help the remote handling process, a cantilever assisting system is suggested to increase the reliability of the transfer operation between vacuum vessel and cask.

  3. Inverse simulation system for evaluating handling qualities during rendezvous and docking

    Science.gov (United States)

    Zhou, Wanmeng; Wang, Hua; Thomson, Douglas; Tang, Guojin; Zhang, Fan

    2017-08-01

    The traditional method used for handling qualities assessment of manned space vehicles is too time-consuming to meet the requirements of an increasingly fast design process. In this study, a rendezvous and docking inverse simulation system to assess the handling qualities of spacecraft is proposed using a previously developed model-predictive-control architecture. By considering the fixed discrete force of the thrusters of the system, the inverse model is constructed using the least squares estimation method with a hyper-ellipsoidal restriction, the continuous control outputs of which are subsequently dispersed by pulse width modulation with sensitivity factors introduced. The inputs in every step are deemed constant parameters, and the method could be considered as a general method for solving nominal, redundant, and insufficient inverse problems. The rendezvous and docking inverse simulation is applied to a nine-degrees-of-freedom platform, and a novel handling qualities evaluation scheme is established according to the operation precision and astronauts' workload. Finally, different nominal trajectories are scored by the inverse simulation and an established evaluation scheme. The scores can offer theoretical guidance for astronaut training and more complex operation missions.

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

    Directory of Open Access Journals (Sweden)

    Giovanni Anthony

    Full Text Available 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.

  5. Effect of push handle height on net moments and forces on the musculoskeletal system during standardized wheelchair pushing tasks

    NARCIS (Netherlands)

    Van Der Woude, L H; Van Koningsbruggen, C M; Kroes, A L; Kingma, I

    1995-01-01

    The aim of this investigation was to analyze the external forces and biomechanical loading on the musculoskeletal system during wheelchair pushing, in relation to different push handle heights. In addition, recommendations for wheelchair pushing in accordance with push handle height are made. Eight

  6. Development of Pneumatic Transport System (PTS) for safe handling of uranium oxide powder in UMP/UED

    International Nuclear Information System (INIS)

    Manna, S.; Satpati, S.K.; Roy, S.B.

    2009-01-01

    Tonnage quantity radioactive uranium oxide powder of particle size sub micron to 100 micron is handled in Uranium Metal Plant (UMP), UED/BARC for production of nuclear grade uranium metal, required for fuelling research reactors - Dhruva and Cirus. A Pneumatic Transfer System (PTS) using vacuum has been introduced and is being used for handling radioactive powder to improve radiation protection

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

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-11-03

    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

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

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2000-01-01

    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

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

    OpenAIRE

    Covarrubias, Mario; Bordegoni, Monica; Cugini, Umberto

    2013-01-01

    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...

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

    International Nuclear Information System (INIS)

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

    1986-12-01

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

  11. Handling And Safety Aspects Of Fiber Optic Laser Beam Delivery Systems

    Science.gov (United States)

    Schonborn, K.-H.; Wodrich, W.

    1988-06-01

    Using lasers for therapeutic applications is getting more and more accepted. In ophthalmology Ar-lasers for intraocular applications are quite common. The Nd:YAG-laser is used as a high power tool in connection with silica fibers for different extracorporal and intracorporal applications. The CO2-laser is the cutting laser, one problem being the beam transmission: The state of the art in fibers is not sufficient up to now. Because of the high power used safety against laser radiation hazard is of great importance. The safety in laser use is primarily dependent on the surgeons cautiousness, e.g. using laser protection goggels, observing that the spot of the aiming beam is present etc. On the other hand the laser and fiber system has to be inherently safe by appropriate technical means as far as possible. An additional aspect adding to safety is the handling: With easier system handling less attention of the surgeon is necessary for driving the apparatus. Thus he can concentrate on the patient and on the procedure. In considering the fiber system one important point in handling and safety is the coupling of the fiber to the laser head. The development philosophy in this coupling may be divided into two groups: - one is trying to use standard connectors which were initially developed for data transmission; - the other is using special connectors. One example of the first group is the guiding of the laser beam from the Ar-laser to the slit-lamp in ophtalmology. Here the well-known F-SMA connectors together with a special fiber with adapted numerical aperture are used. The advantage of such a system is the low price of the connector. For high power lasers such as the clinical Nd:YAG lasers with 40 to 150 W those connectors are not suitable. Up to now every laser manufacturer developed his own connector system in this field.

  12. Monitored Retrievable Storage conceptual system study: dry receiving and handling facility

    International Nuclear Information System (INIS)

    1984-01-01

    A preconceptual design and estimate for a MRS receiving and handling (R and H) facility at a hypothetical site in the United States are presented. The facility consists of a receiving and handling building plus associated operating buildings, system, and site development features. The R and H building and the supporting buildings and site development features are referred to as the R and H area. Adjoining the R and H area will be an interim waste storage area currently being considered by others. The desirability of building a full capacity (3000-MTU) MRS facility initially versus adding additional capacity at a later date in a phased construction program was investigated. Several advantages of phased construction include incorporation of new designs, modification of receiving-handling-packaging, and changes in regulatory requirements or the waste management program which may develop following startup and operation of an 1800-MTU MRS facility. The cost of a 3000-MTU MRS facility constructed initially was estimated at $193,200,000. If a phased construction program was implemented, including escalation to the mid-point of Phase 2 construction, a capital expenditure of $215,300,000 is estimated - a cost penalty of $22,100,000 or about 11% for phased construction

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

    International Nuclear Information System (INIS)

    Sokolov, M. M.; Igonkina, G. B.; Koutcherenko, I. Yu.; Nurov, D. N.

    2008-01-01

    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

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

    International Nuclear Information System (INIS)

    Srinivasan, R.; Sisley, S.E.; Hopf, J.E.

    2004-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, R.; Sisley, S.E.; Hopf, J.E. [BNFL Fuel Solutions, Campbell, CA (United States)

    2004-07-01

    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.

  16. 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.

  17. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2003-02-12

    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. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed.

  18. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2003-01-01

    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. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed

  19. The sample handling system for the Mars Icebreaker Life mission: from dirt to data.

    Science.gov (United States)

    Davé, Arwen; Thompson, Sarah J; McKay, Christopher P; Stoker, Carol R; Zacny, Kris; Paulsen, Gale; Mellerowicz, Bolek; Glass, Brian J; Willson, David; Bonaccorsi, Rosalba; Rask, Jon

    2013-04-01

    The Mars Icebreaker Life mission will search for subsurface life on Mars. It consists of three payload elements: a drill to retrieve soil samples from approximately 1 m below the surface, a robotic sample handling system to deliver the sample from the drill to the instruments, and the instruments themselves. This paper will discuss the robotic sample handling system. Collecting samples from ice-rich soils on Mars in search of life presents two challenges: protection of that icy soil--considered a "special region" with respect to planetary protection--from contamination from Earth, and delivery of the icy, sticky soil to spacecraft instruments. We present a sampling device that meets these challenges. We built a prototype system and tested it at martian pressure, drilling into ice-cemented soil, collecting cuttings, and transferring them to the inlet port of the SOLID2 life-detection instrument. The tests successfully demonstrated that the Icebreaker drill, sample handling system, and life-detection instrument can collectively operate in these conditions and produce science data that can be delivered via telemetry--from dirt to data. Our results also demonstrate the feasibility of using an air gap to prevent forward contamination. We define a set of six analog soils for testing over a range of soil cohesion, from loose sand to basalt soil, with angles of repose of 27° and 39°, respectively. Particle size is a key determinant of jamming of mechanical parts by soil particles. Jamming occurs when the clearance between moving parts is equal in size to the most common particle size or equal to three of these particles together. Three particles acting together tend to form bridges and lead to clogging. Our experiments show that rotary-hammer action of the Icebreaker drill influences the particle size, typically reducing particle size by ≈ 100 μm.

  20. Canister storage building natural phenomena design loads

    International Nuclear Information System (INIS)

    Tallman, A.M.

    1996-02-01

    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

  1. Canister transfer into repository in shaft alternative

    International Nuclear Information System (INIS)

    Raiko, H.; Kukkola, T.; Autio, J.

    2005-09-01

    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.)

  2. Design premises for canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Werme, L

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel 43 refs, 4 figs, 6 tabs

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

    International Nuclear Information System (INIS)

    Robinson, C.; Harris, E.G.; Dyer, P.C.; Williams, J.G.B.

    1985-01-01

    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)

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

    International Nuclear Information System (INIS)

    Sibois, Romain; Määttä, Timo; Siuko, Mikko; Mattila, Jouni

    2014-01-01

    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

  5. Grain boundary corrosion of copper canister material

    International Nuclear Information System (INIS)

    Fennell, P.A.H.; Graham, A.J.; Smart, N.R.; Sofield, C.J.

    2001-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister. The potential for grain boundary corrosion was investigated by exposing copper specimens, which had undergone different heat treatments and hence had different grain sizes, to aerated artificial bentonite-equilibrated groundwater with two concentrations of chloride, for increasing periods of time. The degree of grain boundary corrosion was determined by atomic force microscopy (AFM) and optical microscopy. AFM showed no increase in grain boundary 'ditching' for low chloride groundwater. In high chloride groundwater the surface was covered uniformly with a fine-grained oxide. No increases in oxide thickness were observed. No significant grain boundary attack was observed using optical microscopy either. The work suggests that in aerated artificial groundwaters containing chloride ions, grain boundary corrosion of copper is unlikely to adversely affect SKB's copper canisters

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

    International Nuclear Information System (INIS)

    Vere, B.; Mathevon, P.

    1985-01-01

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

  7. Methodology and results of risk assessment of interconnections within the JET active gas handling system

    International Nuclear Information System (INIS)

    Ballantyne, P.R.; Bell, A.C.; Konstantellos, A.; Hemmerich, J.L.

    1992-01-01

    The Joint European Torus (JET) Active Gas Handling System (AGHS) is a complex interconnection of numerous subsystems. While individual subsystems were assessed for their risk of operation, an assessment of the effects of inadvertent interconnections was needed. A systematic method to document the assessment was devised to ease the assessment of complex plant and was applied to the AGHS. The methodology, application to AGHS, the four critical issues and required plant modifications as a result of this assessment are briefly discussed in this paper

  8. Am/Cm canister temperature evaluation in CIM5

    International Nuclear Information System (INIS)

    Baich, M.A.

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

    Dana, W.P.

    1995-12-01

    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

  10. 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.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  13. Remotely controlled inspection and handling systems for decommissioning tasks in nuclear facilities

    International Nuclear Information System (INIS)

    Schreck, G.; Bach, W.; Haferkamp, H.

    1993-01-01

    The Institut fur Werkstoffkunde at the University of Hanover has recently developed three remotely controlled systems for different underwater inspection and dismantling tasks. ODIN I is a tool guiding device, particularly being designed for the dismantling of the steam dryer housing of the KRB A power plant at Gundremmingen, Germany. After being approved by the licencing organization TUEV Bayern, hot operation started in November 1992. The seven axes remotely controlled handling system ZEUS, consisting of a three translatory axes guiding machine and a tool handling device with four rotatory axes, has been developed for the demonstration of underwater plasma arc cutting of spherical metallic components with great wall thicknesses. A specially designed twin sensor system and a modular torch, exchanged by means of a remote controlled tool changing device, will be used for different complex cutting tasks. FAUST, an autonomous, freediving underwater vehicle, was designed for complex inspection, maintenance and dismantling tasks. It is equipped with two video cameras, an ultrasonic and a radiologic sensor and a small plasma torch. A gripper and a subsidiary vehicle for inspection may be attached. (author)

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

    International Nuclear Information System (INIS)

    Teese, G.D.; Randall, W.J.

    1992-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  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. 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.

  19. 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

    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 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 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...

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

    International Nuclear Information System (INIS)

    Santaliz, Jorge O.; Paredes, Juan A.

    1998-01-01

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

  1. Settlement of Canisters with smectite clay envelopes in deposition holes

    International Nuclear Information System (INIS)

    Pusch, R.

    1986-12-01

    Settlement of canisters containing radioactive waste and being surrounded by dense smectite clay is caused by the stresses and heat induced in the clay. Consolidation by water expulsion of the clay underlying a model canister with 5 cm diameter and 30 cm length would theoretically account for a maximum finite settlement of about 70 my m in a few weeks, while shear-induced creep would yield a settlement of only a few microns in the same time period. These predictions were checked by running a laboratory test in which a dead load of 80 kg was applied to a small cylindrical copper canister embedded in Na bentonite. The settlement, which increased in proportion to log time, turned out to be about 6 my m in the first 2.5 months. After the first loading period at room temperature, heating to 50 degrees C and, after a 4 months long 'room temperature' period, to 70 degrees C took place. This cycling gave strong, instant settlement and upheaval because of the different thermal expansion of the interacting components of the system. After the development of constant temperature conditions in the entire system and completion of the consolidation or expansion that followed from the thermo-mechanical interactions, the settlement proceeded at a rather high rate at 70 degrees C, still following a log time creep law, but with somewhat stronger retardation. At room temperature, i.e. in the post-heating periods, the settlement seemed to cease, on the other hand. The conclusion from the study is that the canister movements under isothermal conditions were in accordance with the log t-type creep settlement that was predicted in theoretical grounds. Pre-heating and low stresses may account for extraordinary retardation of the settlement. (author)

  2. Closed-system drug-transfer devices plus safe handling of hazardous drugs versus safe handling alone for reducing exposure to infusional hazardous drugs in healthcare staff.

    Science.gov (United States)

    Gurusamy, Kurinchi Selvan; Best, Lawrence Mj; Tanguay, Cynthia; Lennan, Elaine; Korva, Mika; Bussières, Jean-François

    2018-03-27

    Occupational exposure to hazardous drugs can decrease fertility and result in miscarriages, stillbirths, and cancers in healthcare staff. Several recommended practices aim to reduce this exposure, including protective clothing, gloves, and biological safety cabinets ('safe handling'). There is significant uncertainty as to whether using closed-system drug-transfer devices (CSTD) in addition to safe handling decreases the contamination and risk of staff exposure to infusional hazardous drugs compared to safe handling alone. To assess the effects of closed-system drug-transfer of infusional hazardous drugs plus safe handling versus safe handling alone for reducing staff exposure to infusional hazardous drugs and risk of staff contamination. We searched the Cochrane Central Register of Controlled Trials (CENTRAL), MEDLINE, Embase, OSH-UPDATE, CINAHL, Science Citation Index Expanded, economic evaluation databases, the World Health Organization International Clinical Trials Registry Platform, and ClinicalTrials.gov to October 2017. We included comparative studies of any study design (irrespective of language, blinding, or publication status) that compared CSTD plus safe handling versus safe handling alone for infusional hazardous drugs. Two review authors independently identified trials and extracted data. We calculated the risk ratio (RR) and mean difference (MD) with 95% confidence intervals (CI) using both fixed-effect and random-effects models. We assessed risk of bias according to the risk of bias in non-randomised studies of interventions (ROBINS-I) tool, used an intracluster correlation coefficient of 0.10, and we assessed the quality of the evidence using GRADE. We included 23 observational cluster studies (358 hospitals) in this review. We did not find any randomised controlled trials or formal economic evaluations. In 21 studies, the people who used the intervention (CSTD plus safe handling) and control (safe handling alone) were pharmacists or pharmacy

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

    International Nuclear Information System (INIS)

    Kalbantner, P.

    2001-08-01

    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

  4. Topical safety analysis report for the transportation of the NUHOMS reg-sign dry shielded canister

    International Nuclear Information System (INIS)

    1993-08-01

    This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS reg-sign) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS reg-sign DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS reg-sign Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport

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

    International Nuclear Information System (INIS)

    Griffiths, J.S.

    1985-10-01

    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

  6. Efficiency Evaluation of Handling of Geologic-Geophysical Information by Means of Computer Systems

    Science.gov (United States)

    Nuriyahmetova, S. M.; Demyanova, O. V.; Zabirova, L. M.; Gataullin, I. I.; Fathutdinova, O. A.; Kaptelinina, E. A.

    2018-05-01

    Development of oil and gas resources, considering difficult geological, geographical and economic conditions, requires considerable finance costs; therefore their careful reasons, application of the most perspective directions and modern technologies from the point of view of cost efficiency of planned activities are necessary. For ensuring high precision of regional and local forecasts and modeling of reservoirs of fields of hydrocarbonic raw materials, it is necessary to analyze huge arrays of the distributed information which is constantly changing spatial. The solution of this task requires application of modern remote methods of a research of the perspective oil-and-gas territories, complex use of materials remote, nondestructive the environment of geologic-geophysical and space methods of sounding of Earth and the most perfect technologies of their handling. In the article, the authors considered experience of handling of geologic-geophysical information by means of computer systems by the Russian and foreign companies. Conclusions that the multidimensional analysis of geologicgeophysical information space, effective planning and monitoring of exploration works requires broad use of geoinformation technologies as one of the most perspective directions in achievement of high profitability of an oil and gas industry are drawn.

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

    Energy Technology Data Exchange (ETDEWEB)

    Garzoglio, Gabriele [DePaul Univ., Chicago, IL (United States)

    2006-01-13

    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.

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

    International Nuclear Information System (INIS)

    Tuominen, Janne; Rasi, Teemu; Mattila, Jouni; Siuko, Mikko; Esque, Salvador; Hamilton, David

    2013-01-01

    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

  9. 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.

  10. 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.

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

    International Nuclear Information System (INIS)

    Han Zandong; Luo Sheng; Liu Jiguo

    2008-01-01

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

  12. 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.

  13. Decontamination of stainless steel canisters that contain high-level waste

    International Nuclear Information System (INIS)

    Bray, L.A.

    1987-01-01

    At the West Valley Nuclear Services Company (WVNSC) in West Valley, New York, high-level radioactive waste (HLW) will be vitrified into a borosilicate glass form and poured into large, stainless steel canisters. During the filling process, volatile fission products, principally 137 Cs, condense on the exterior of the canisters. The smearable contamination that remains on the canisters after they are filled and partially cooled must be removed from the canisters' exterior surfaces prior to their storage and ultimate shipment to a US Department of Energy (DOE) repository for disposal. A simple and effective method was developed for decontamination of HLW canisters. This method of chemical decontamination is applicable to a wide variety of contaminated equipment found in the nuclear industry. The process employs a reduction-oxidation system [Ce(III)/Ce(IV)] in nitric acid solution to chemically mill the surface of stainless steel, similar to the electropolishing process, but without the need for an applied electrical current. Contaminated canisters are simply immersed in the solution at controlled temperature and Ce(IV) concentration levels

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. Deep Borehole Disposal Concept: Development of Universal Canister Concept of Operations

    Energy Technology Data Exchange (ETDEWEB)

    Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Applied Systems Analysis and Research; Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Applied Systems Analysis and Research

    2016-08-01

    This report documents key elements of the conceptual design for deep borehole disposal of radioactive waste to support the development of a universal canister concept of operations. A universal canister is a canister that is designed to be able to store, transport, and dispose of radioactive waste without the canister having to be reopened to treat or repackage the waste. This report focuses on the conceptual design for disposal of radioactive waste contained in a universal canister in a deep borehole. The general deep borehole disposal concept consists of drilling a borehole into crystalline basement rock to a depth of about 5 km, emplacing WPs in the lower 2 km of the borehole, and sealing and plugging the upper 3 km. Research and development programs for deep borehole disposal have been ongoing for several years in the United States and the United Kingdom; these studies have shown that deep borehole disposal of radioactive waste could be safe, cost effective, and technically feasible. The design concepts described in this report are workable solutions based on expert judgment, and are intended to guide follow-on design activities. Both preclosure and postclosure safety were considered in the development of the reference design concept. The requirements and assumptions that form the basis for the deep borehole disposal concept include WP performance requirements, radiological protection requirements, surface handling and transport requirements, and emplacement requirements. The key features of the reference disposal concept include borehole drilling and construction concepts, WP designs, and waste handling and emplacement concepts. These features are supported by engineering analyses.

  16. 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....

  17. Robust, Radiation Tolerant Command and Data Handling and Power System Electronics for SmallSats

    Science.gov (United States)

    Nguyen, Hanson Cao; Fraction, James

    2018-01-01

    In today's budgetary environment, there is significant interest within the National Aeronautics and Space Administration (NASA) to enable small robotic science missions that can be executed faster and cheaper than previous larger missions. To help achieve this, focus has shifted from using exclusively radiation-tolerant or radiation-hardened parts to using more commercial-off-the-shelf (COTS) components for NASA small satellite missions that can last at least one year in orbit. However, there are some portions of a spacecraft's avionics, such as the Command and Data Handling (C&DH) subsystem and the Power System Electronics (PSE) that need to have a higher level of reliability that goes beyond what is attainable with currently available COTS parts. While there are a number of COTS components that can withstand a total ionizing dose (TID) of tens or hundreds of kilorads, there is still a great deal of concern about tolerance to and mitigation of single-event effects (SEE).

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

    International Nuclear Information System (INIS)

    Staples, A.T.

    1991-01-01

    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)

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

    International Nuclear Information System (INIS)

    Nascimento Filho, V.F. do; Barros Ferraz, E.S. de

    1986-01-01

    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) [pt

  20. Overview of the performance of the JET active gas handling system during and after DTE1

    International Nuclear Information System (INIS)

    Laesser, R.; Atkins, G.; Bell, A.

    1999-02-01

    The JET Active Gas Handling System (AGHS) was designed, built and commissioned to handle safely radioactive tritium gas mixtures, to supply tritium (T 2 ) and deuterium (D 2 ) to the JET torus, to process the exhaust gases with the main purpose to enrich and re-use T 2 and D 2 , to detritiate tritiated impurities and to keep discharges far below the approved daily release limits. In addition, the AGHS had to supply the necessary ventilation air streams during maintenance or repair inside or outside of the AGHS building. During the first Deuterium-Tritium Experiment (DTE1) at JET in 1997 the AGHS fulfilled all these tasks in an excellent manner. No unauthorised or unplanned tritium releases occurred and no operational delays were caused by the AGHS. In fact, this was the first true demonstration that quantities of tritium in the tens of grams range can be processed and recycled safely and efficiently in a large fusion device. At the start of DTE1 20 g of tritium were available on the JET site. About 100 g of tritium were supplied from the AGHS to the users which necessitated the recycling of tritium at least five times. Approximately 220 tritium plasma shots were performed during DTE1. Large amounts of tritium were temporarily trapped in the torus. This overview presents the performance of the whole AGHS during DTE1 as well as general aspects such as the preparation for DTE1; the quantities of gases supplied from the AGHS to the users and pumped back to the AGHS; tritium accountancy; interlock systems; failure of equipment; and gives detailed information of the gas processing in each subsystem of the AGHS. As a consequence of the performance of the AGHS during DTE1 we can state confidently that the AGHS is ready for further Deuterium-Tritium Experiments. (author)

  1. Immune system handling time may alter the outcome of competition between pathogens and the immune system.

    Science.gov (United States)

    Greenspoon, Philip B; Banton, Sydney; Mideo, Nicole

    2018-06-14

    Predators may be limited in their ability to kill prey (i.e., have type II or III functional responses), an insight that has had far-reaching consequences in the ecological literature. With few exceptions, however, this possibility has not been extended to the behaviour of immune cells, which kill pathogens much as predators kill their prey. Rather, models of the within-host environment have tended to tacitly assume that immune cells have an unlimited ability to target and kill pathogens (i.e., a type I functional response). Here we explore the effects of changing this assumption on infection outcomes (i.e., pathogen loads). We incorporate immune cell handling time into an ecological model of the within-host environment that considers both the predatory nature of the pathogen-immune cell interaction as well as competition between immune cells and pathogens for host resources. Unless pathogens can preempt immune cells for host resources, adding an immune cell handling time increases equilibrium pathogen load. We find that the shape of the relationship between energy intake and pathogen load can change: with a type I functional response, pathogen load is maximised at intermediate inputs, while for a type II or III functional response, pathogen load is solely increasing. With a type II functional response, pathogen load can fluctuate rather than settling to an equilibrium, a phenomenon unobserved with type I or III functional responses. Our work adds to a growing literature highlighting the role of resource availability in host-parasite interactions. Implications of our results for adaptive anorexia are discussed. Copyright © 2018 Elsevier Ltd. All rights reserved.

  2. Engineering Support for Handling Controller Conflicts in Energy Storage Systems Applications

    Directory of Open Access Journals (Sweden)

    Claudia Zanabria

    2017-10-01

    Full Text Available Energy storage systems will play a major role in the decarbonization of future sustainable electric power systems, allowing a high penetration of distributed renewable energy sources and contributing to the distribution network stability and reliability. To accomplish this, a storage system is required to provide multiple services such as self-consumption, grid support, peak-shaving, etc. The simultaneous activation of controllers operation may lead to conflicts, as a consequence the execution of committed services is not guaranteed. This paper presents and discusses a solution to the exposed issue by developing an engineering support approach to semi-automatically detect and handle conflicts for multi-usage storage systems applications. To accomplish that an ontology is developed and exploited by model-driven engineering mechanisms. The proposed approach is evaluated by implementing a use case example, where detection of conflicts is automatically done at an early design stage. Besides this, exploitable source code for conflicts resolution is generated and used during the design and prototype stages of controllers development. Thus, the proposed engineering support enhances the design and development of storage system controllers, especially for multi-usage applications.

  3. Memory handling in the ATLAS submission system from job definition to sites limits

    CERN Document Server

    Forti, Alessandra; The ATLAS collaboration

    2016-01-01

    The ATLAS workload management system is a pilot system based on a late binding philosophy that avoided for many years to pass fine grained job requirements to the batch system. In particular for memory most of the requirements were set to request 4GB vmem as defined in the EGI portal VO card, i.e. 2GB RAM + 2GB swap. However in the past few years several changes have happened in the operating system kernel and in the applications that make such a definition of memory to use for requesting slots obsolete and ATLAS has introduced the new PRODSYS2 workload management which has a more flexible system to evaluate the memory requirements and to submit to appropriate queues. The work stemmed in particular from the introduction of 64bit multicore workloads and the increased memory requirements of some of the single core applications. This paper describes the overall review and changes of memory handling starting from the definition of tasks, the way tasks memory requirements are set using scout jobs and the new memor...

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

    International Nuclear Information System (INIS)

    Boergesson, Lennart; Hernelind, Jan

    2006-02-01

    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/m 3 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/m 3 . 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Boergesson, Lennart [Clay Technology AB, Lund (Sweden); Hernelind, Jan [FEMTech AB, Vaesteraas (Sweden)

    2006-02-15

    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/m{sup 3} 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/m{sup 3}. 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

  6. 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...

  7. Memory handling in the ATLAS submission system from job definition to sites limits

    Science.gov (United States)

    Forti, A. C.; Walker, R.; Maeno, T.; Love, P.; Rauschmayr, N.; Filipcic, A.; Di Girolamo, A.

    2017-10-01

    In the past few years the increased luminosity of the LHC, changes in the linux kernel and a move to a 64bit architecture have affected the ATLAS jobs memory usage and the ATLAS workload management system had to be adapted to be more flexible and pass memory parameters to the batch systems, which in the past wasn’t a necessity. This paper describes the steps required to add the capability to better handle memory requirements, included the review of how each component definition and parametrization of the memory is mapped to the other components, and what changes had to be applied to make the submission chain work. These changes go from the definition of tasks and the way tasks memory requirements are set using scout jobs, through the new memory tool developed to do that, to how these values are used by the submission component of the system and how the jobs are treated by the sites through the CEs, batch systems and ultimately the kernel.

  8. The European contribution to the procurement of the ITER Remote Handling systems

    International Nuclear Information System (INIS)

    Damiani, Carlo; Irving, Mike; Semeraro, Luigi

    2009-01-01

    Fusion for Energy (F4E) will manage the European in-kind contribution of various remote handling (RH) systems for the maintenance of ITER components: (i) the divertor cassette movers, end effectors, manipulator arms and tooling; (ii) 50% of the transfer casks, in particular the air transfer systems and some in-cask devices; (iii) the in-vessel viewing and metrology system (IVVS); (iv) the Neutral Beam (NB) Cell crane, manipulator arms, tooling, Caesium Oven replacement tooling, NB source installation/removal trolley, auxiliary vehicles. A wide range of technologies is involved: special monorail crane, movers, manipulator arms, pipe cutting/welding tooling, special cameras, laser-based metrology devices, control systems, virtual reality. An important aspect to consider is the resistance to radiation levels that range from max ∼10 KGy/h for IVVS down to ∼1 Gy/h for the RH devices operating in the NB cell. Given the unprecedented complexity of the ITER maintenance scenario, a development strategy is being implemented that includes prototyping and testing of RH subsystems before proceeding with the final production for ITER. This paper presents an overview of the various procurement packages, the status of development for each of them, the validation and procurement strategy, including issues like radiation resistance and standardisation policy, and the organisational and managerial challenges in relation with the complex ITER Organisation (IO).

  9. Memory handling in the ATLAS submission system from job definition to sites limits

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00027700; The ATLAS collaboration; Walker, Rodney; Maeno, Tadashi; Love, Peter; Rauschmayr, Nathalie; Filipcic, Andrej; Di Girolamo, Alessandro

    2017-01-01

    In the past few years the increased luminosity of the LHC, changes in the linux kernel and a move to a 64bit architecture have affected the ATLAS jobs memory usage and the ATLAS workload management system had to be adapted to be more flexible and pass memory parameters to the batch systems, which in the past wasn’t a necessity. This paper describes the steps required to add the capability to better handle memory requirements, included the review of how each component definition and parametrization of the memory is mapped to the other components, and what changes had to be applied to make the submission chain work. These changes go from the definition of tasks and the way tasks memory requirements are set using scout jobs, through the new memory tool developed to do that, to how these values are used by the submission component of the system and how the jobs are treated by the sites through the CEs, batch systems and ultimately the kernel.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  11. 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.

  12. Monitored retrievable storage and multi-purpose canister robotic applications: Feasibility, dose savings and cost analysis

    International Nuclear Information System (INIS)

    Bennett, P.C.

    1995-01-01

    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, and as an alternative to current MPC closure and welding methods at commercial nuclear reactor sites. Automation of key operational aspects is analyzed to determine equipment requirements, through-put times and equipment costs. The economic analysis approach is described, and economic and radiation dose impacts resulting from this automation are compared to manual handling methods. (author). 5 refs, 5 figs, 3 tabs

  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)

    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

  15. Copper canisters for nuclear high level waste disposal. Corrosion aspects

    International Nuclear Information System (INIS)

    Werme, L.; Sellin, P.; Kjellbert, N.

    1992-10-01

    A corrosion analysis of a thick-walled copper canister for spent fuel disposal is discussed. The analysis has shown that there are no rapid mechanisms that may lead to canister failure, indicating an anticipated corrosion service life of several millions years. If further analysis of the copper canister is considered, it should be concentrated on identifying and evaluating processes other than corrosion, which may have a potential for leading to canister failure. (au)

  16. Multi-purpose canister storage unit and transfer cask thermal analysis

    International Nuclear Information System (INIS)

    Montgomery, R.A.; Niemer, K.A.; Lindner, C.N.

    1997-01-01

    Spent Nuclear Fuel (SNF) generated at commercial nuclear power plants throughout the US is a concern because of continued delays in obtaining a safe, permanent disposal facility. Most utilities maintain their SNF in wet storage pools; however, after decades of use, many pools are filled to capacity. Unfortunately, DOE's proposed final repository at Yucca Mountain is at least 10 years from completion, and commercial power utilities have few options for SNF storage in the interim. The Multi-Purpose Canister (MPC) system, sponsored by DOE's Office of Civilian Radioactive Waste Management, is a viable solution to the interim storage problem. The system is designed for interim dry storage, transport, and ultimate disposal of commercial SNF. The MPC system consists of four separate components: an MPC, Transfer Cask, Storage Unit, and Transport Cask. The SNF assemblies are loaded and sealed inside the helium-filled steel MPC. Once sealed, the MPC is not reopened, eliminating the need to re-handle the individual spent fuel assemblies. The MPC is transferred, using the MPC Transfer Cask, into a cylindrical, reinforced-concrete Storage Unit for on-site dry storage. The MPC may be removed from the Storage Unit at any time and transferred into the MPC Transport Cask for transport to the final repository. This paper discusses the analytical approach used to evaluate the heat transfer characteristics of an MPC containing SNF assemblies in the MPC Transfer Cask and Storage Unit

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

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    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

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

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    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

  19. An overview of process instrumentation, protective safety interlocks and alarm system at the JET facilities active gas handling system

    International Nuclear Information System (INIS)

    Skinner, N.; Brennan, P.; Brown, K.; Gibbons, C.; Jones, G.; Knipe, S.; Manning, C.; Perevezentsev, A.; Stagg, R.; Thomas, R.; Yorkshades, J.

    2003-01-01

    The Joint European Torus (JET) Facilities Active Gas Handling System (AGHS) comprises ten interconnected processing sub-systems that supply, process and recover tritium from gases used in the JET Machine. Operations require a diverse range of process instrumentation to carry out a multiplicity of monitoring and control tasks and approximately 500 process variables are measured. The different types and application of process instruments are presented with specially adapted or custom-built versions highlighted. Forming part of the Safety Case for tritium operations, a dedicated hardwired interlock and alarm system provides an essential safety function. In the event of failure modes, each hardwired interlock will back-up software interlocks and shutdown areas of plant to a failsafe condition. Design of the interlock and alarm system is outlined and general methodology described. Practical experience gained during plant operations is summarised and the methods employed for routine functional testing of essential instrument systems explained

  20. Material handling for the Los Alamos National Laboratory Nuclear Storage Facility

    International Nuclear Information System (INIS)

    Pittman, P.; Roybal, J.; Durrer, R.; Gordon, D.

    1999-01-01

    This paper will present the design and application of material handling and automation systems currently being developed for the Los Alamos National Laboratory (LANL) Nuclear Material Storage Facility (NMSF) renovation project. The NMSF is a long-term storage facility for nuclear material in various forms. The material is stored within tubes in a rack called a basket. The material handling equipment range from simple lift assist devices to more sophisticated fully automated robots, and are split into three basic systems: a Vault Automation System, an NDA automation System, and a Drum handling System. The Vault Automation system provides a mechanism to handle a basket of material cans and to load/unload storage tubes within the material vault. In addition, another robot is provided to load/unload material cans within the baskets. The NDA Automation System provides a mechanism to move material within the small canister NDA laboratory and to load/unload the NDA instruments. The Drum Handling System consists of a series of off the shelf components used to assist in lifting heavy objects such as pallets of material or drums and barrels

  1. 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.

  2. Force sensitive handles and capacitive touch sensor for driving a flexible haptic-based immersive system.

    Science.gov (United States)

    Covarrubias, Mario; Bordegoni, Monica; Cugini, Umberto

    2013-10-09

    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.

  3. Thermo-hydro-mechanical mode of canister retrieval test

    International Nuclear Information System (INIS)

    Zandarin, M.T.; Olivella, S.; Gens', A.; Alonso, E.E.

    2010-01-01

    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

  4. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    Werme, L.

    1990-09-01

    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.)

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

    International Nuclear Information System (INIS)

    Grundmann, J.G.

    1974-01-01

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

  6. Neutron-sensitive ZnS/10B2O3 ceramic scintillator detector as an alternative to a 3He-gas-based detector for a plutonium canister assay system

    International Nuclear Information System (INIS)

    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.

    2014-01-01

    A neutron-sensitive ZnS/ 10 B 2 O 3 ceramic scintillator detector was developed as an alternative to a 3 He-gas-based detector for use in a plutonium canister assay system. The detector has a modular structure, with a flat ZnS/ 10 B 2 O 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 137 Cs 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

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

    International Nuclear Information System (INIS)

    Kring, C.T.; Herndon, J.N.; Meacham, S.A.

    1987-01-01

    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

  8. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    McCoy, H.E.

    1983-01-01

    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 300 0 C, 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 300 0 C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 800 0 C 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

  9. Plutonium Immobilization Project - Robotic canister loading

    International Nuclear Information System (INIS)

    Hamilton, R.L.

    2000-01-01

    The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site (SRS), Lawrence Livermore National Laboratory (LLNL), Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL). When operational in 2008, the PIP will fulfill the nation's nonproliferation commitment by placing surplus weapons-grade plutonium in a permanently stable ceramic form and making it unattractive for reuse. Since there are significant radiation and security concerns, the program team is developing novel and unique technology to remotely perform plutonium immobilization tasks. The remote task covered in this paper employs a jointed arm robot to load seven 3.5 inch diameter, 135-pound cylinders (magazines) through the 4 inch diameter neck of a stainless steel canister. Working through the narrow canister neck, the robot secures the magazines into a specially designed rack pre-installed in the canister. To provide the deterrent effect, the canisters are filled with a mixture of high-level waste and glass at the Defense Waste Processing Facility (DWPF)

  10. Decontamination of high-level waste canisters

    International Nuclear Information System (INIS)

    Nesbitt, J.F.; Slate, S.C.; Fetrow, L.K.

    1980-12-01

    This report presents evaluations of several methods for the in-process decontamination of metallic canisters containing any one of a number of solidified high-level waste (HLW) forms. The use of steam-water, steam, abrasive blasting, electropolishing, liquid honing, vibratory finishing and soaking have been tested or evaluated as potential techniques to decontaminate the outer surfaces of HLW canisters. Either these techniques have been tested or available literature has been examined to assess their applicability to the decontamination of HLW canisters. Electropolishing has been found to be the most thorough method to remove radionuclides and other foreign material that may be deposited on or in the outer surface of a canister during any of the HLW processes. Steam or steam-water spraying techniques may be adequate for some applications but fail to remove all contaminated forms that could be present in some of the HLW processes. Liquid honing and abrasive blasting remove contamination and foreign material very quickly and effectively from small areas and components although these blasting techniques tend to disperse the material removed from the cleaned surfaces. Vibratory finishing is very capable of removing the bulk of contamination and foreign matter from a variety of materials. However, special vibratory finishing equipment would have to be designed and adapted for a remote process. Soaking techniques take long periods of time and may not remove all of the smearable contamination. If soaking involves pickling baths that use corrosive agents, these agents may cause erosion of grain boundaries that results in rough surfaces

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

    International Nuclear Information System (INIS)

    Datta, D.; Hegde, A.G.

    2001-01-01

    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)

  12. 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

    2016-05-19

    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.

  13. 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.

  14. 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

  15. Dual Source Time-of-flight Mass Spectrometer and Sample Handling System

    Science.gov (United States)

    Brinckerhoff, W.; Mahaffy, P.; Cornish, T.; Cheng, A.; Gorevan, S.; Niemann, H.; Harpold, D.; Rafeek, S.; Yucht, D.

    We present details of an instrument under development for potential NASA missions to planets and small bodies. The instrument comprises a dual ionization source (laser and electron impact) time-of-flight mass spectrometer (TOF-MS) and a carousel sam- ple handling system for in situ analysis of solid materials acquired by, e.g., a coring drill. This DSTOF instrument could be deployed on a fixed lander or a rover, and has an open design that would accommodate measurements by additional instruments. The sample handling system (SHS) is based on a multi-well carousel, originally de- signed for Champollion/DS4. Solid samples, in the form of drill cores or as loose chips or fines, are inserted through an access port, sealed in vacuum, and transported around the carousel to a pyrolysis cell and/or directly to the TOF-MS inlet. Samples at the TOF-MS inlet are xy-addressable for laser or optical microprobe. Cups may be ejected from their holders for analyzing multiple samples or caching them for return. Samples are analyzed with laser desorption and evolved-gas/electron-impact sources. The dual ion source permits studies of elemental, isotopic, and molecular composition of unprepared samples with a single mass spectrometer. Pulsed laser desorption per- mits the measurement of abundance and isotope ratios of refractory elements, as well as the detection of high-mass organic molecules in solid samples. Evolved gas analysis permits similar measurements of the more volatile species in solids and aerosols. The TOF-MS is based on previous miniature prototypes at JHU/APL that feature high sensitivity and a wide mass range. The laser mode, in which the sample cup is directly below the TOF-MS inlet, permits both ablation and desorption measurements, to cover elemental and molecular species, respectively. In the evolved gas mode, sample cups are raised into a small pyrolysis cell and heated, producing a neutral gas that is elec- tron ionized and pulsed into the TOF-MS. (Any imaging

  16. Conceptual design report for a remotely operated cask handling system. Revision 1

    International Nuclear Information System (INIS)

    Yount, J.A.; Berger, J.D.

    1984-09-01

    Recent advances in remote handling utilizing commercial robotics are conceptually applied to lowering operator cumulative radiation exposure and increasing throughput during cask handling operations in nuclear shipping and receiving facilities. Revision 1 incorporates functional criteria for facility equipment, equipment technical outline specifications, and interface control drawings to assist Architect Engineers in the application of remote handling to waste shipping and receiving facilities. The document has also been updated to show some of the equipment used in proof-of-principle testing during fiscal year 1984. 10 references, 50 figures, 1 table

  17. Techniques for remote maintenance of in-cell material-handling system in the HFEF/N main cell

    International Nuclear Information System (INIS)

    Tobias, D.A.; Frickey, C.A.

    1975-01-01

    Operations in the main cell of HFEF/N have required development of remote handling equipment and unique techniques for maintaining the in-cell material-handling system. Specially designed equipment is used to remove a disabled crane or electromechanical manipulator bridge from its support rails and place it on floor stands for repair or maintenance. Support areas for the main cell, such as the spray chamber and hot repair area, provide essential decontamination, repair, and staging areas for the in-cell material-handling-system equipment and tools. A combined engineering and technical effort in upgrading existing master-slave manipulators has definitely reduced the requirements for their maintenance. The cell is primarily for postirradiation examination of LMFBR materials and fuel elements

  18. PENINGKATAN KINERJA OHMS (ORDER HANDLING MANUFACTURING SYSTEM) MELALUI SOFT SYSTEMS METHODOLOGY (SSM)

    OpenAIRE

    Slamet Budiarto

    2011-01-01

    Dalam penelitian ini, untuk meningkatkan kinerja OHMS yang karakteristiknya lebih kompleks, digunakan soft systems methodology yang banyak membantu para manajer untuk menyelesaikan masalah yang bervariasi dan kompleks. Permasalahan tersebut sering menemui kegagalan dalam penyelesaiannya ketika pendekatan system engineering (SE) digunakan. Penelitian menunjukan gambaran aktivitas lebih jelas, dan penentuan indikator lebih terstruktur dengan mengkombinasikan IDEF0. Melalui soft systems met...

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  20. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    1996-12-01

    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.)

  1. Criticality safety calculations for the nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    Anttila, M.

    1996-12-01

    The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent fuel has been studied with the MCNP4A code based on the Monte Carlo technique and with the fuel assembly burnup programs CASMO-HEX and CASMO-4. Two rather similar types of spent fuel disposal canisters have been studied. One canister type has been designed for hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant (IVO canister) and the other one for square BWR fuel bundles used at the Olkiluoto nuclear power plant (TVO canister). (10 refs.)

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

    International Nuclear Information System (INIS)

    KESSLER, S.F.

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

    NELSON, J.V.

    1999-01-01

    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

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

    International Nuclear Information System (INIS)

    Dytz, Aline Guerra; Dullius, Marcos Alexandre; Gomes, Camila e Silva

    2008-01-01

    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)

  5. 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.

  6. A dedicated database system for handling multi-level data in systems biology.

    Science.gov (United States)

    Pornputtapong, Natapol; Wanichthanarak, Kwanjeera; Nilsson, Avlant; Nookaew, Intawat; Nielsen, Jens

    2014-01-01

    Advances in high-throughput technologies have enabled extensive generation of multi-level omics data. These data are crucial for systems biology research, though they are complex, heterogeneous, highly dynamic, incomplete and distributed among public databases. This leads to difficulties in data accessibility and often results in errors when data are merged and integrated from varied resources. Therefore, integration and management of systems biological data remain very challenging. To overcome this, we designed and developed a dedicated database system that can serve and solve the vital issues in data management and hereby facilitate data integration, modeling and analysis in systems biology within a sole database. In addition, a yeast data repository was implemented as an integrated database environment which is operated by the database system. Two applications were implemented to demonstrate extensibility and utilization of the system. Both illustrate how the user can access the database via the web query function and implemented scripts. These scripts are specific for two sample cases: 1) Detecting the pheromone pathway in protein interaction networks; and 2) Finding metabolic reactions regulated by Snf1 kinase. In this study we present the design of database system which offers an extensible environment to efficiently capture the majority of biological entities and relations encountered in systems biology. Critical functions and control processes were designed and implemented to ensure consistent, efficient, secure and reliable transactions. The two sample cases on the yeast integrated data clearly demonstrate the value of a sole database environment for systems biology research.

  7. Canister disposition plan for the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.; Payne, C.H.

    1990-01-01

    This report details the disposition of canisters and the canistered waste forms produced during the DWPF Startup Test Program. The six melter campaigns (DWPF Startup Tests FA-13, WP-14, WP-15, WP-16, WP-17, and FA-18) will produce 126 canistered waste forms. In addition, up to 20 additional canistered waste forms may be produced from glass poured during the transition between campaigns. In particular, this canister disposition plan (1) assigns (by alpha-numeric code) a specific canister to each location in the six campaign sequences, (2) describes the method of access for glass sampling on each canistered waste form, (3) describes the nature of the specific tests which will be carried out, (4) details which tests will be carried out on each canistered waste form, (5) provides the sequence of these tests for each canistered waste form, and (6) assigns a storage location for each canistered waste form. The tests are designed to provide evidence, as detailed in the Waste Form Compliance Plan (WCP 1 ), that the DWPF product will comply with the Waste Acceptance Product Specifications (WAPS 2 ). The WAPS must be met before the canistered waste form is accepted by DOE for ultimate disposal at the Federal Repository. The results of these tests will be included in the Waste Form Qualification Report (WQR)

  8. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    Werme, L.

    1991-12-01

    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

  9. 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.

  10. Design report of the disposal canister for twelve fuel assemblies

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    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.)

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

    International Nuclear Information System (INIS)

    Ziaei, Z.; Hahto, A.; Mattila, J.; Siuko, M.; Semeraro, L.

    2011-01-01

    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.

  12. 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.

  13. Development and use of a remote waste handling system for disposal of greater confinement wastes

    International Nuclear Information System (INIS)

    Williams, R.E.

    1985-01-01

    This paper discusses the design and development of a remotely controlled waste handling system (RWHS) for use in radioactive waste disposal operations. A RWHS was developed at the US Department of Energy's (DOE) Nevada Test Site for use in the Greater Confinement Disposal Test (GCDT). The RWHS consists of a remote control console and the following remotely operated features: a crane, a grapple/manipulator module which is suspended by the crane hoist hook, and closed-circuit television cameras. The RWHS was used to safely place high-specific-activity radioactive waste in greater confinement disposal. Between December 15, 1983, and February 23, 1984, five encapsulated sources were open-air transferred from shielded shipping casks and placed 30 m down a 3-m-dia augered shaft using the RWHS. These sources contained approximately 460 kCi of 90 Sr, 21 kCi of 137 Cs, and 390 Ci of 60 Co. Each source was transferred safely and efficiently and operational personnel did not receive any recordable doses. 3 references, 5 figures

  14. Implications of the 'Energide' concept for communication and information handling in the central nervous system.

    Science.gov (United States)

    Agnati, L F; Fuxe, K; Baluska, F; Guidolin, D

    2009-08-01

    Recently a revision of the cell theory has been proposed, which has several implications both for physiology and pathology. This revision is founded on adapting the old Julius von Sach's proposal (1892) of the Energide as the fundamental universal unit of eukaryotic life. This view maintains that, in most instances, the living unit is the symbiotic assemblage of the cell periphery complex organized around the plasma membrane, some peripheral semi-autonomous cytosol organelles (as mitochondria and plastids, which may be or not be present), and of the Energide (formed by the nucleus, microtubules, and other satellite structures). A fundamental aspect is the proposal that the Energide plays a pivotal and organizing role of the entire symbiotic assemblage (see Appendix 1). The present paper discusses how the Energide paradigm implies a revision of the concept of the internal milieu. As a matter of fact, the Energide interacts with the cytoplasm that, in turn, interacts with the interstitial fluid, and hence with the medium that has been, classically, known as the internal milieu. Some implications of this aspect have been also presented with the help of a computational model in a mathematical Appendix 2 to the paper. Finally, relevances of the Energide concept for the information handling in the central nervous system are discussed especially in relation to the inter-Energide exchange of information.

  15. Proposal for Construction/Demonstration/Implementation of A Material Handling System

    International Nuclear Information System (INIS)

    Jim Jnatt