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Sample records for container system description

  1. 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

  2. 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

  3. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

    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. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials

  4. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Uncanistered Spent Nuclear Fuel (SNF) 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 with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident

  5. System Description:

    DEFF Research Database (Denmark)

    Schürmann, Carsten; Poswolsky, Adam

    2009-01-01

    Delphin is a functional programming language [Adam Poswolsky and Carsten Schürmann. Practical programming with higher-order encodings and dependent types. In European Symposium on Programming (ESOP), 2008] utilizing dependent higher-order datatypes. Delphin's two-level type-system cleanly separates...

  6. TMACS system description

    International Nuclear Information System (INIS)

    Scaief, C.C.

    1995-01-01

    This document provides a description of the Tank Monitor and Control System (TMACS). It is intended as an introduction for those persons unfamiliar with the system as well as a reference document for the users, maintenance personnel, and system designers. In addition to describing the system, the document outlines the associated drawing documentation, provides maintenance and spare parts information, and discusses other TMACS documents that provide additional detail

  7. Management control system description

    Energy Technology Data Exchange (ETDEWEB)

    Bence, P. J.

    1990-10-01

    This Management Control System (MCS) description describes the processes used to manage the cost and schedule of work performed by Westinghouse Hanford Company (Westinghouse Hanford) for the US Department of Energy, Richland Operations Office (DOE-RL), Richland, Washington. Westinghouse Hanford will maintain and use formal cost and schedule management control systems, as presented in this document, in performing work for the DOE-RL. This MCS description is a controlled document and will be modified or updated as required. This document must be approved by the DOE-RL; thereafter, any significant change will require DOE-RL concurrence. Westinghouse Hanford is the DOE-RL operations and engineering contractor at the Hanford Site. Activities associated with this contract (DE-AC06-87RL10930) include operating existing plant facilities, managing defined projects and programs, and planning future enhancements. This document is designed to comply with Section I-13 of the contract by providing a description of Westinghouse Hanford's cost and schedule control systems used in managing the above activities. 5 refs., 22 figs., 1 tab.

  8. Integrated Project Management System description

    International Nuclear Information System (INIS)

    1987-03-01

    The Uranium Mill Tailings Remedial Action (UMTRA) Project is a Department of Energy (DOE) designated Major System Acquisition (MSA). To execute and manage the Project mission successfully and to comply with the MSA requirements, the UMTRA Project Office (''Project Office'') has implemented and operates an Integrated Project Management System (IPMS). The Project Office is assisted by the Technical Assistance Contractor's (TAC) Project Integration and Control (PIC) Group in system operation. Each participant, in turn, provides critical input to system operation and reporting requirements. The IPMS provides a uniform structured approach for integrating the work of Project participants. It serves as a tool for planning and control, workload management, performance measurement, and specialized reporting within a standardized format. This system description presents the guidance for its operation. Appendices 1 and 2 contain definitions of commonly used terms and abbreviations and acronyms, respectively. 17 figs., 5 tabs

  9. ELECTRICAL SUPPORT SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    S. Roy

    2004-06-24

    The purpose of this revision of the System Design Description (SDD) is to establish requirements that drive the design of the electrical support system and their bases to allow the design effort to proceed to License Application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience/users are design engineers. This type of SDD both ''leads'' and ''trails'' the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. Functional and operational requirements applicable to electrical support systems are obtained from the ''Project Functional and Operational Requirements'' (F&OR) (Siddoway 2003). Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (Doraswamy 2004), and fire hazards analyses. The above-mentioned low-level documents address ''Project Requirements Document'' (PRD) (Canon and Leitner 2003) requirements. This SDD contains several appendices that include supporting information. Appendix B lists key system charts, diagrams, drawings, and lists, and Appendix C includes a list of system procedures.

  10. SNF AGING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    L.L. Swanson

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F andOR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts, diagrams, drawings, lists and additional supporting information; and Appendix C is a list of

  11. ELECTRICAL SUPPORT SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    Roy, S.

    2004-01-01

    The purpose of this revision of the System Design Description (SDD) is to establish requirements that drive the design of the electrical support system and their bases to allow the design effort to proceed to License Application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience/users are design engineers. This type of SDD both ''leads'' and ''trails'' the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. Functional and operational requirements applicable to electrical support systems are obtained from the ''Project Functional and Operational Requirements'' (F andOR) (Siddoway 2003). Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (Doraswamy 2004), and fire hazards analyses. The above-mentioned low-level documents address ''Project Requirements Document'' (PRD) (Canon and Leitner 2003) requirements. This SDD contains several appendices that include supporting information. Appendix B lists key system charts, diagrams, drawings, and lists, and Appendix C includes a list of system procedures

  12. The transportation operations system: A description

    International Nuclear Information System (INIS)

    Best, R.E.; Danese, F.L.; Dixon, L.D.; Peterson, R.W.; Pope, R.B.

    1990-01-01

    This paper presents a description of the system for transporting radioactive waste that may be deployed to accomplish the assigned system mission, which includes accepting spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from waste generator sites and transporting them to the FWMS destination facilities. The system description presented here contains, in part, irradiated fuel and waste casks, ancillary equipments, truck, rail, and barge transporters, cask and vehicle traffic management organizations, maintenance facilities, and other operations elements. The description is for a fully implemented system, which is not expected to be achieved, however, until several years after initial operations. 6 figs

  13. SNF AGING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    L.L. Swanson

    2005-04-06

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F&OR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts

  14. Passive containment system

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1977-01-01

    Disclosed is a containment system that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is prevented from over-heating by a high containment back-pressure and a reactor vessel refill system. The primary containment vessel is restored to a high sub-atmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means. 20 claims, 14 figures

  15. Containment vessel drain system

    Science.gov (United States)

    Harris, Scott G.

    2018-01-30

    A system for draining a containment vessel may include a drain inlet located in a lower portion of the containment vessel. The containment vessel may be at least partially filled with a liquid, and the drain inlet may be located below a surface of the liquid. The system may further comprise an inlet located in an upper portion of the containment vessel. The inlet may be configured to insert pressurized gas into the containment vessel to form a pressurized region above the surface of the liquid, and the pressurized region may operate to apply a surface pressure that forces the liquid into the drain inlet. Additionally, a fluid separation device may be operatively connected to the drain inlet. The fluid separation device may be configured to separate the liquid from the pressurized gas that enters the drain inlet after the surface of the liquid falls below the drain inlet.

  16. Epsilon. A System Description Language

    DEFF Research Database (Denmark)

    Jensen, Kurt; Kyng, Morten

    This paper discusses the use of Petri nets as a semantic tool in the design of languages and in the construction and analysis of system descriptions. The topics treated are: -- Languages based on nets. -- The problem of time in nets. -- Nets and related models. -- Nets and formal semantics...

  17. Description of the containment for a stationary pressurized water reactor

    International Nuclear Information System (INIS)

    Hermani, S.

    1986-01-01

    The function of the containment is to prevent the inadvertent release of radioactive fission products from the reactor coolant system to the atmosphere and to provide biological shielding during both normal and accident operation. Basically three different containment concepts 1) the dry containment, 2) the subatmospheric containment, and 3) the ice condenser containment, have been developed, based on how the accident energy release from the reactor coolant system is controlled. The containment structure can be either 1) reinforced concrets with inside liner, 2) prestressed concrete with inside, or 3) full steel cylinder or steel sphere with separate concrete shield. The size of the containment is largely dictated by the required net free volume, that satisfies the energy release criteria due to the design basic accident. The design and construction methods applied to this structure guarantee that the containment will carry out its safety function. This was proven by the Three Mile Island accident. (author)

  18. Advanced Containment System

    Science.gov (United States)

    Kostelnik, Kevin M.; Kawamura, Hideki; Richardson, John G.; Noda, Masaru

    2004-10-12

    An advanced containment system for containing buried waste and associated leachate. A trench is dug on either side of the zone of interest containing the buried waste so as to accommodate a micro tunnel boring machine. A series of small diameter tunnels are serially excavated underneath the buried waste. The tunnels are excavated by the micro tunnel boring machine at a consistent depth and are substantially parallel to each other. As tunneling progresses, steel casing sections are connected end to end in the excavated portion of the tunnel so that a steel tube is formed. Each casing section has complementary interlocking structure running its length that interlocks with complementary interlocking structure on the adjacent casing section. Thus, once the first tube is emplaced, placement of subsequent tubes is facilitated by the complementary interlocking structure on the adjacent, previously placed, casing sections.

  19. Containment heat removal system

    International Nuclear Information System (INIS)

    Wade, G.E.; Barbanti, G.; Gou, P.F.; Rao, A.S.; Hsu, L.C.

    1992-01-01

    This patent describes a nuclear system of a type including a containment having a nuclear reactor therein, the nuclear reactor including a pressure vessel and a core in the pressure vessel, the system. It comprises a gravity pool of coolant disposed at an elevation sufficient to permit a flow of coolant into the nuclear reactor pressure vessel against a predetermined pressure within the nuclear reactor pressure vessel; means for reducing a pressure of steam in the nuclear reactor pressure vessel to a value less than the predetermined pressure in the event of a nuclear accident, the means including a depressurization valve connected to the pressure vessel, the means further including steam heat dissipating means such dissipating means including a suppression pool; a supply of water in the suppression pool, there being a headspace in the suppression pool above the water supply; a substantial amount of air in the head space; means for feeding pressurized steam from the nuclear reactor pressure vessel to a location under a surface of the supply of water, the supply of water being effective to absorb heat sufficient to reduce steam pressure below the predetermined pressure; and a check valve for communicating the headspace with the containment, the check valve being oriented to vent air in the headspace to the containment when a pressure in the headspace exceeds a pressure in the containment by a predetermined pressure differential

  20. 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

  1. 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

  2. ELECTRICAL POWER SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    M. Maniyar

    2004-06-22

    The purpose of this revision of the System Description Document (SDD) is to establish requirements that drive the design of the electrical power system and their bases to allow the design effort to proceed to License Application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience are design engineers. This type of SDD leads and follows the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential to performing the design process. This SDD follows the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. Functional and operational requirements applicable to this system are obtained from ''Project Functional and Operational Requirements'' (F&OR) (Siddoway, 2003). Other requirements to support the design process have been taken from higher level requirements documents such as ''Project Design Criteria Document'' (PDC) (Doraswamy 2004), the fire hazards analyses, and the preclosure safety analysis. The above mentioned low-level documents address ''Project Requirements Document'' (PRD) (Canori and Leitner 2003) requirements. This SDD includes several appendices with supporting information. Appendix B lists key system charts, diagrams, drawings, and lists; and Appendix C is a list of system procedures.

  3. Data Description of a System

    Directory of Open Access Journals (Sweden)

    P. Nevriva

    1996-04-01

    Full Text Available In this paper, a brief discussion on description of process by memorized data is given. The insight into the problem can offer modified views on optimal control, on data compression at communication systems with respect to information content of message, etc.The idea of process description by memorized data with different information content will be presented here on the classical case study of optimal control: the data based control algorithm (data algorithm, DA gathers data from the controlled process and derives control signal (control from data accumulated in the data base. The implementation of the DA on the ideal computer which is not limited by its speed or capacity of memory is expected for simplicity. Accuracy of the data algorithm is then given by a-priori knowledge of the task and by information exchange between the controlled process and the computer.

  4. ASSEMBLY TRANSFER SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    Gorpani, B.

    2000-01-01

    The Assembly Transfer System (ATS) receives, cools, and opens rail and truck transportation casks from the Carrier/Cask Handling System (CCHS). The system unloads transportation casks consisting of bare Spent Nuclear Fuel (SNF) assemblies, single element canisters, and Dual Purpose Canisters (DPCs). For casks containing DPCs, the system opens the DPCs and unloads the SNF. The system stages the assemblies, transfer assemblies to and from fuel-blending inventory pools, loads them into Disposal Containers (DCs), temporarily seals and inerts the DC, decontaminates the DC and transfers it to the Disposal Container Handling System. The system also prepares empty casks and DPCs for off-site shipment. Two identical Assembly Transfer System lines are provided in the Waste Handling Building (WHB). Each line operates independently to handle the waste transfer throughput and to support maintenance operations. Each system line primarily consists of wet and dry handling areas. The wet handling area includes a cask transport system, cask and DPC preparation system, and a wet assembly handling system. The basket transport system forms the transition between the wet and dry handling areas. The dry handling area includes the dry assembly handling system, assembly drying system, DC preparation system, and DC transport system. Both the wet and dry handling areas are controlled by the control and tracking system. The system operating sequence begins with moving transportation casks to the cask preparation area. The cask preparation operations consist of cask cavity gas sampling, cask venting, cask cool-down, outer lid removal, and inner shield plug lifting fixture attachment. Casks containing bare SNF (no DPC) are filled with water and placed in the cask unloading pool. The inner shield plugs are removed underwater. For casks containing a DPC, the cask lid(s) is removed, and the DPC is penetrated, sampled, vented, and cooled. A DPC lifting fixture is attached and the cask is placed

  5. Content of system design descriptions

    International Nuclear Information System (INIS)

    1998-10-01

    A System Design Description (SDD) describes the requirements and features of a system. This standard provides guidance on the expected technical content of SDDs. The need for such a standard was recognized during efforts to develop SDDs for safety systems at DOE Hazard Category 2 nonreactor nuclear facilities. Existing guidance related to the corresponding documents in other industries is generally not suitable to meet the needs of DOE nuclear facilities. Across the DOE complex, different contractors have guidance documents, but they vary widely from site to site. While such guidance documents are valuable, no single guidance document has all the attributes that DOE considers important, including a reasonable degree of consistency or standardization. This standard is a consolidation of the best of the existing guidance. This standard has been developed with a technical content and level of detail intended to be most applicable to safety systems at DOE Hazard Category 2 nonreactor nuclear facilities. Notwithstanding that primary intent, this standard is recommended for other systems at such facilities, especially those that are important to achieving the programmatic mission of the facility. In addition, application of this standard should be considered for systems at other facilities, including non-nuclear facilities, on the basis that SDDs may be beneficial and cost-effective

  6. Integrated Project Management System description

    International Nuclear Information System (INIS)

    1994-09-01

    The Integrated Program Management System (IPMS) Description is a ''working'' document that describes the work processes of the Uranium Mill Tailings Remedial Action Project Office (UMTRA) and IPMS Group. This document has undergone many revisions since the UMTRA Project began; this revision not only updates the work processes but more clearly explains the relationships between the Project Office, contractors, and other participants. The work process flow style has been revised to better describe Project work and the relationships of participants. For each work process, more background and guidance on ''why'' and ''what is expected'' is given. For example, a description of activity data sheets has been added in the work organization and the Project performance and reporting processes, as well as additional detail about the federal budget process and funding management and improved flow charts and explanations of cost and schedule management. A chapter has been added describing the Cost Reduction/Productivity Improvement Program. The Change Control Board (CCB) procedures (Appendix A) have been updated. Project critical issues meeting (PCIM) procedures have been added as Appendix B. Budget risk assessment meeting procedures have been added as Appendix C. These appendices are written to act as stand-alone documentation for each process. As the procedures are improved and updated, the documentation can be updated separately

  7. Subatmospheric double containment system

    International Nuclear Information System (INIS)

    Gans, D. Jr.; Noble, J.H.

    1978-01-01

    A reinforced concrete double wall nuclear containment structure with each wall including an essentially impervious membrane or liner and porous concrete filling the annulus between the two walls is described. The interior of the structure is maintained at subatmospheric pressure, and the annulus between the two walls is maintained at a subatmospheric pressure intermediate between that of the interior and the surrounding atmospheric pressure, during normal operation. In the event of an accident within the containment structure the interior pressure may exceed atmospheric pressure, but leakage from the interior to the annulus between the double walls will not result in the pressure of the annulus exceeding atmospheric pressure so that there is no net outleakage from the containment structure

  8. Nuclear steam system containment

    International Nuclear Information System (INIS)

    Jabsen, F.

    1980-01-01

    An improved containment used for radiation shielding and pressure suppression comprising a dry well includes a pressure vessel, a plurality of concentric wall means, said plurality of concentric wall means defining at least three annular regions about said dry well. A first annular region provides the containment used for radiation shielding, a second annular region is substantially dry, a third annular region provides a wet well for relieving fluid pressure released from the pressure vessel into the dry well. Pipe connection means extend in the wet well from the dry well, a pool of liquid is disposed to partially fill said third annular region, the upper end portion of the second and third annular regions having an enclosure, and a plurality of baffle plates extending vertically downward from said enclosure in said third annular region into said pool of liquid so as to circumferentially divide the upper portion of said third annular region into a plurality of circumferential upper portions

  9. Description of an improvement concept to prevent overpressure containment rupture

    International Nuclear Information System (INIS)

    Covelli, B.

    1985-01-01

    This report summarizes results of experiments and recommendations for design improvements, shown by the example of a standard PWR-type system designed in Western Germany. The design improvements are intended to allow safe handling of the hydrogen problem and prevention of undue pressure built-up in the containment. Dimensions and design data are given of the technical components in order to present a realistic view of the measures to be taken for accident prevention. The measures described have been tested and proved to afford optimal advances with regard to prevention of a hydrogen explosion, by inerting with Halon; controlled venting, by means of an open filtering system with head-end blow-off condenser; after-heat removal, by an appropriately dimensioned blow-off condenser, or by means of an additional external spray cooling system. (orig./HP) [de

  10. Seal containment system

    International Nuclear Information System (INIS)

    Kugler, R.W.; Gerkey, K.S.; Kasner, W.H.

    1978-01-01

    An automated system for transporting nuclear fuel elements between fuel element assembly stations without contaminating the area outside the sealed assembly stations is described. The system comprises a plurality of assembly stations connected together by an elongated horizontal sealing mechanism and an automatic transport mechanism for transporting a nuclear fuel element in a horizontal attitude between the assembly stations while the open end of the fuel element extends through the sealing mechanism into the assembly station enclosure. The sealing mechanism allows the fuel element to be advanced by the transport mechanism while limiting the escape of radioactive particles from within the assembly station enclosure. 4 claims, 6 figures

  11. DOE-RL Integrated Safety Management System Description

    International Nuclear Information System (INIS)

    SHOOP, D.S.

    2000-01-01

    The purpose of this Integrated Safety Management System Description (ISMSD) is to describe the U.S. Department of Energy (DOE), Richland Operations Office (RL) ISMS as implemented through the RL Integrated Management System (RIMS). This ISMSD does not impose additional requirements but rather provides an overview describing how various parts of the ISMS fit together. Specific requirements for each of the core functions and guiding principles are established in other implementing processes, procedures, and program descriptions that comprise RIMS. RL is organized to conduct work through operating contracts; therefore, it is extremely difficult to provide an adequate ISMS description that only addresses RL functions. Of necessity, this ISMSD contains some information on contractor processes and procedures which then require RL approval or oversight. This ISMSD does not purport to contain a full description of the contractors' ISM System Descriptions

  12. DOE-RL Integrated Safety Management System Description

    CERN Document Server

    Shoop, D S

    2000-01-01

    The purpose of this Integrated Safety Management System Description (ISMSD) is to describe the U.S. Department of Energy (DOE), Richland Operations Office (RL) ISMS as implemented through the RL Integrated Management System (RIMS). This ISMSD does not impose additional requirements but rather provides an overview describing how various parts of the ISMS fit together. Specific requirements for each of the core functions and guiding principles are established in other implementing processes, procedures, and program descriptions that comprise RIMS. RL is organized to conduct work through operating contracts; therefore, it is extremely difficult to provide an adequate ISMS description that only addresses RL functions. Of necessity, this ISMSD contains some information on contractor processes and procedures which then require RL approval or oversight. This ISMSD does not purport to contain a full description of the contractors' ISM System Descriptions.

  13. Hot Spot Removal System: System description

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    Hazardous wastes contaminated with radionuclides, chemicals, and explosives exist across the Department of Energy complex and need to be remediated due to environmental concerns. Currently, an opportunity is being developed to dramatically reduce remediation costs and to assist in the acceleration of schedules associated with these wastes by deploying a Hot Spot Removal System. Removing the hot spot from the waste site will remove risk driver(s) and enable another, more cost effective process/option/remedial alternative (i.e., capping) to be applied to the remainder of the site. The Hot Spot Removal System consists of a suite of technologies that will be utilized to locate and remove source terms. Components of the system can also be used in a variety of other cleanup activities. This Hot Spot Removal System Description document presents technologies that were considered for possible inclusion in the Hot Spot Removal System, technologies made available to the Hot Spot Removal System, industrial interest in the Hot Spot Removal System`s subsystems, the schedule required for the Hot Spot Removal System, the evaluation of the relevant technologies, and the recommendations for equipment and technologies as stated in the Plan section.

  14. Hot Spot Removal System: System description

    International Nuclear Information System (INIS)

    1997-09-01

    Hazardous wastes contaminated with radionuclides, chemicals, and explosives exist across the Department of Energy complex and need to be remediated due to environmental concerns. Currently, an opportunity is being developed to dramatically reduce remediation costs and to assist in the acceleration of schedules associated with these wastes by deploying a Hot Spot Removal System. Removing the hot spot from the waste site will remove risk driver(s) and enable another, more cost effective process/option/remedial alternative (i.e., capping) to be applied to the remainder of the site. The Hot Spot Removal System consists of a suite of technologies that will be utilized to locate and remove source terms. Components of the system can also be used in a variety of other cleanup activities. This Hot Spot Removal System Description document presents technologies that were considered for possible inclusion in the Hot Spot Removal System, technologies made available to the Hot Spot Removal System, industrial interest in the Hot Spot Removal System''s subsystems, the schedule required for the Hot Spot Removal System, the evaluation of the relevant technologies, and the recommendations for equipment and technologies as stated in the Plan section

  15. 1995 Baseline solid waste management system description

    International Nuclear Information System (INIS)

    Anderson, G.S.; Konynenbelt, H.S.

    1995-09-01

    This provides a detailed solid waste system description that documents the treatment, storage, and disposal (TSD) strategy for managing Hanford's solid low-level waste, low-level mixed waste, transuranic and transuranic mixed waste, and greater-than-Class III waste. This system description is intended for use by managers of the solid waste program, facility and system planners, as well as system modelers. The system description identifies the TSD facilities that constitute the solid waste system and defines these facilities' interfaces, schedules, and capacities. It also provides the strategy for treating each of the waste streams generated or received by the Hanford Site from generation or receipt through final destination

  16. Waste Management System Description Document (WMSD)

    International Nuclear Information System (INIS)

    1992-02-01

    This report is an appendix of the ''Waste Management Description Project, Revision 1''. This appendix is about the interim approach for the technical baseline of the waste management system. It describes the documentation and regulations of the waste management system requirements and description. (MB)

  17. Technical description of the CANDU nuclear power system

    International Nuclear Information System (INIS)

    Husseini, S.D.

    1977-01-01

    The lecture consists of: 1.) KANUPP description: general - Containment- Reactor Cooling System - Moderator Helium System. 2.) Operating experience: Plant Performances (as applicable to Kanupp) - Major failures of primary system - Performance of primary circulating pumps. - Heavy Water Leakage Control. - Radiation Dose Control - Inadvertant Operations. (orig.) [de

  18. Authenticated Secure Container System (ASCS)

    International Nuclear Information System (INIS)

    1991-01-01

    Sandia National Laboratories developed an Authenticated Secure Container System (ASCS) for the International Atomic Energy Agency (IAEA). Agency standard weights and safeguards samples can be stored in the ASCS to provide continuity of knowledge. The ASCS consists of an optically clear cover, a base containing the Authenticated Item Monitoring System (AIMS) transmitter, and the AIMS receiver unit for data collection. The ASCS will provide the Inspector with information concerning the status of the system, during a surveillance period, such as state of health, tampering attempts, and movement of the container system. The secure container is located inside a Glove Box with the receiver located remotely from the Glove Box. AIMS technology uses rf transmission from the secure container to the receiver to provide a record of state of health and tampering. The data is stored in the receiver for analysis by the Inspector during a future inspection visit. 2 refs

  19. System Design Description for the TMAD Code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

    This document serves as the System Design Description (SDD) for the TMAD Code System, which includes the TMAD code and the LIBMAKR code. The SDD provides a detailed description of the theory behind the code, and the implementation of that theory. It is essential for anyone who is attempting to review or modify the code or who otherwise needs to understand the internal workings of the code. In addition, this document includes, in Appendix A, the System Requirements Specification for the TMAD System

  20. Subsurface Ventilation System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    Eric Loros

    2001-07-25

    The Subsurface Ventilation System supports the construction and operation of the subsurface repository by providing air for personnel and equipment and temperature control for the underground areas. Although the system is located underground, some equipment and features may be housed or located above ground. The system ventilates the underground by providing ambient air from the surface throughout the subsurface development and emplacement areas. The system provides fresh air for a safe work environment and supports potential retrieval operations by ventilating and cooling emplacement drifts. The system maintains compliance within the limits established for approved air quality standards. The system maintains separate ventilation between the development and waste emplacement areas. The system shall remove a portion of the heat generated by the waste packages during preclosure to support thermal goals. The system provides temperature control by reducing drift temperature to support potential retrieval operations. The ventilation system has the capability to ventilate selected drifts during emplacement and retrieval operations. The Subsurface Facility System is the main interface with the Subsurface Ventilation System. The location of the ducting, seals, filters, fans, emplacement doors, regulators, and electronic controls are within the envelope created by the Ground Control System in the Subsurface Facility System. The Subsurface Ventilation System also interfaces with the Subsurface Electrical System for power, the Monitored Geologic Repository Operations Monitoring and Control System to ensure proper and safe operation, the Safeguards and Security System for access to the emplacement drifts, the Subsurface Fire Protection System for fire safety, the Emplacement Drift System for repository performance, and the Backfill Emplacement and Subsurface Excavation Systems to support ventilation needs.

  1. Subsurface Ventilation System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-12

    The Subsurface Ventilation System supports the construction and operation of the subsurface repository by providing air for personnel and equipment and temperature control for the underground areas. Although the system is located underground, some equipment and features may be housed or located above ground. The system ventilates the underground by providing ambient air from the surface throughout the subsurface development and emplacement areas. The system provides fresh air for a safe work environment and supports potential retrieval operations by ventilating and cooling emplacement drifts. The system maintains compliance within the limits established for approved air quality standards. The system maintains separate ventilation between the development and waste emplacement areas. The system shall remove a portion of the heat generated by the waste packages during preclosure to support thermal goals. The system provides temperature control by reducing drift temperature to support potential retrieval operations. The ventilation system has the capability to ventilate selected drifts during emplacement and retrieval operations. The Subsurface Facility System is the main interface with the Subsurface Ventilation System. The location of the ducting, seals, filters, fans, emplacement doors, regulators, and electronic controls are within the envelope created by the Ground Control System in the Subsurface Facility System. The Subsurface Ventilation System also interfaces with the Subsurface Electrical System for power, the Monitored Geologic Repository Operations Monitoring and Control System to ensure proper and safe operation, the Safeguards and Security System for access to the emplacement drifts, the Subsurface Fire Protection System for fire safety, the Emplacement Drift System for repository performance, and the Backfill Emplacement and Subsurface Excavation Systems to support ventilation needs.

  2. Ground Control System Description Document

    International Nuclear Information System (INIS)

    Eric Loros

    2001-01-01

    The Ground Control System contributes to the safe construction and operation of the subsurface facility, including accesses and waste emplacement drifts, by maintaining the configuration and stability of the openings during construction, development, emplacement, and caretaker modes for the duration of preclosure repository life. The Ground Control System consists of ground support structures installed within the subsurface excavated openings, any reinforcement made to the rock surrounding the opening, and inverts if designed as an integral part of the system. The Ground Control System maintains stability for the range of geologic conditions expected at the repository and for all expected loading conditions, including in situ rock, construction, operation, thermal, and seismic loads. The system maintains the size and geometry of operating envelopes for all openings, including alcoves, accesses, and emplacement drifts. The system provides for the installation and operation of sensors and equipment for any required inspection and monitoring. In addition, the Ground Control System provides protection against rockfall for all subsurface personnel, equipment, and the engineered barrier system, including the waste package during the preclosure period. The Ground Control System uses materials that are sufficiently maintainable and that retain the necessary engineering properties for the anticipated conditions of the preclosure service life. These materials are also compatible with postclosure waste isolation performance requirements of the repository. The Ground Control System interfaces with the Subsurface Facility System for operating envelopes, drift orientation, and excavated opening dimensions, Emplacement Drift System for material compatibility, Monitored Geologic Repository Operations Monitoring and Control System for ground control instrument readings, Waste Emplacement/Retrieval System to support waste emplacement operations, and the Subsurface Excavation System

  3. Subsurface Facility System Description Document

    International Nuclear Information System (INIS)

    Eric Loros

    2001-01-01

    The Subsurface Facility System encompasses the location, arrangement, size, and spacing of the underground openings. This subsurface system includes accesses, alcoves, and drifts. This system provides access to the underground, provides for the emplacement of waste packages, provides openings to allow safe and secure work conditions, and interfaces with the natural barrier. This system includes what is now the Exploratory Studies Facility. The Subsurface Facility System physical location and general arrangement help support the long-term waste isolation objectives of the repository. The Subsurface Facility System locates the repository openings away from main traces of major faults, away from exposure to erosion, above the probable maximum flood elevation, and above the water table. The general arrangement, size, and spacing of the emplacement drifts support disposal of the entire inventory of waste packages based on the emplacement strategy. The Subsurface Facility System provides access ramps to safely facilitate development and emplacement operations. The Subsurface Facility System supports the development and emplacement operations by providing subsurface space for such systems as ventilation, utilities, safety, monitoring, and transportation

  4. Formal Description of Hybrid Systems

    DEFF Research Database (Denmark)

    Zhou, Chaochen; Ji, Wang; Ravn, Anders P.

    1996-01-01

    A language to describe hybrid systems, i.e. networks of communicating discrete and continuous processes, is proposed. A semantics of the language is given in Extended Duration Calculus, a real-time interval logic with a proof system that allows reasoning in mathematical analysis about continuous ...

  5. System Description for the Double Shell Tank (DST) Confinement System

    International Nuclear Information System (INIS)

    ROSSI, H.

    2000-01-01

    This document provides a description of the Double-Shell Tank (DST) Confinement System. This description will provide a basis for developing functional, performance and test requirements (i.e., subsystem specification), as necessary, for the DST Confinement System

  6. Simulation of containment atmosphere stratification experiment using local instantaneous description

    International Nuclear Information System (INIS)

    Babic, M.; Kljenak, I.

    2004-01-01

    An experiment on mixing and stratification in the atmosphere of a nuclear power plant containment at accident conditions was simulated with the CFD code CFX4.4. The original experiment was performed in the TOSQAN experimental facility. Simulated nonhomogeneous temperature, species concentration and velocity fields are compared to experimental results. (author)

  7. NDMAS System and Process Description

    Energy Technology Data Exchange (ETDEWEB)

    Larry Hull

    2012-10-01

    Experimental data generated by the Very High Temperature Reactor Program need to be more available to users in the form of data tables on Web pages that can be downloaded to Excel or in delimited text formats that can be used directly for input to analysis and simulation codes, statistical packages, and graphics software. One solution that can provide current and future researchers with direct access to the data they need, while complying with records management requirements, is the Nuclear Data Management and Analysis System (NDMAS). This report describes the NDMAS system and its components, defines roles and responsibilities, describes the functions the system performs, describes the internal processes the NDMAS team uses to carry out the mission, and describes the hardware and software used to meet Very High Temperature Reactor Program needs.

  8. Descriptive Analyses of Mechanical Systems

    DEFF Research Database (Denmark)

    Andreasen, Mogens Myrup; Hansen, Claus Thorp

    2003-01-01

    Forord Produktanalyse og teknologianalyse kan gennmføres med et bredt socio-teknisk sigte med henblik på at forstå kulturelle, sociologiske, designmæssige, forretningsmæssige og mange andre forhold. Et delområde heri er systemisk analyse og beskrivelse af produkter og systemer. Nærværende kompend...

  9. 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

  10. Technical description of the Swedish natural gas distribution system

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Ronny [KM Miljoeteknik AB (Sweden)

    1997-06-01

    This description of the Swedish distribution network has been produced to provide information for distribution companies, trade organisations, etc., who have an interest in getting a clear understanding of the technical design and standards, technical directives, etc., which have served as guidance in the development. The technical description covers the piping system from a measuring and regulating station (MR station) up to the consumer`s substation, however, only sections with a maximum operating pressure of 4 bar. By way of introduction, the description contains introductory information on supply channels, consumption patterns and the principal design of the high pressure network in Sweden 10 refs, 10 figs, 1 tab

  11. SURFACE INDUSTRIAL HVAC SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    M.M. Ansari

    2005-04-05

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

  12. Waste Management Systems Requirements and Descriptions (SRD)

    International Nuclear Information System (INIS)

    Conner, C.W.

    1986-01-01

    The Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the development of a system for the management of high-level radioactive waste and spent fuel in accordance with the Nuclear Waste Policy Act of 1982. The Waste Management system requirements and description document is the program-level technical baseline document. The requirements include the functions that must be performed in order to achieve the system mission and performance criteria for those functions. This document covers only the functional requirements of the system; it does not cover programmatic or procedural requirements pertaining to the processes of designing, siting and licensing. The requirements are largely based on the Nuclear Waste Policy Act of 1982, Environmental Protection Agency standards, Nuclear Regulatory Commission regulations, and DOE orders and guidance. However, nothing in this document should be construed as to relieve the DOE or its contractors from their responsibilities to comply with applicable statutes, regulations, and standards. This document also provides a brief description of the system being developed to meet the requirements. In addition to the described ''authorized system,'' a system description is provided for an ''improved-performance system'' which would include a monitored retrievable storage (MRS) facility. In the event that an MRS facility is approved by Congress, the improved-performance system will become the reference system. Neither system description includes Federal Interim Storage (FIS) capabilities. Should the need for FIS be identified, it will be included as an additional system element. The descriptions are focused on the interfaces between the system elements, rather than on the detail of the system elements themselves

  13. Computer software design description for the integrated control and data acquisition system LDUA system

    International Nuclear Information System (INIS)

    Aftanas, B.L.

    1998-01-01

    This Computer Software Design Description (CSDD) document provides the overview of the software design for all the software that is part of the integrated control and data acquisition system of the Light Duty Utility Arm System (LDUA). It describes the major software components and how they interface. It also references the documents that contain the detailed design description of the components

  14. 1993 baseline solid waste management system description

    International Nuclear Information System (INIS)

    Armacost, L.L.; Fowler, R.A.; Konynenbelt, H.S.

    1994-02-01

    Pacific Northwest Laboratory has prepared this report under the direction of Westinghouse Hanford Company. The report provides an integrated description of the system planned for managing Hanford's solid low-level waste, low-level mixed waste, transuranic waste, and transuranic mixed waste. The primary purpose of this document is to illustrate a collective view of the key functions planned at the Hanford Site to handle existing waste inventories, as well as solid wastes that will be generated in the future. By viewing this system as a whole rather than as individual projects, key facility interactions and requirements are identified and a better understanding of the overall system may be gained. The system is described so as to form a basis for modeling the system at various levels of detail. Model results provide insight into issues such as facility capacity requirements, alternative system operating strategies, and impacts of system changes (ie., startup dates). This description of the planned Hanford solid waste processing system: defines a baseline system configuration; identifies the entering waste streams to be managed within the system; identifies basic system functions and waste flows; and highlights system constraints. This system description will evolve and be revised as issues are resolved, planning decisions are made, additional data are collected, and assumptions are tested and changed. Out of necessity, this document will also be revised and updated so that a documented system description, which reflects current system planning, is always available for use by engineers and managers. It does not provide any results generated from the many alternatives that will be modeled in the course of analyzing solid waste disposal options; such results will be provided in separate documents

  15. Advanced Transport Operating System (ATOPS) control display unit software description

    Science.gov (United States)

    Slominski, Christopher J.; Parks, Mark A.; Debure, Kelly R.; Heaphy, William J.

    1992-01-01

    The software created for the Control Display Units (CDUs), used for the Advanced Transport Operating Systems (ATOPS) project, on the Transport Systems Research Vehicle (TSRV) is described. Module descriptions are presented in a standardized format which contains module purpose, calling sequence, a detailed description, and global references. The global reference section includes subroutines, functions, and common variables referenced by a particular module. The CDUs, one for the pilot and one for the copilot, are used for flight management purposes. Operations performed with the CDU affects the aircraft's guidance, navigation, and display software.

  16. System description of the Repository-Only System for the FY 1990 systems integration program studies

    International Nuclear Information System (INIS)

    McKee, R.W.; Young, J.R.; Konzek, G.J.

    1991-07-01

    This document provides both functional and physical descriptions of a conceptual high-level waste management system defined as a Repository-Only System. Its purpose is to provide a basis for required system computer modeling and system studies initiated in FY 1990 under the Systems Integration Program of the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The Repository-Only System is designed to accept 3000 MTU per year of spent fuel and 400 equivalent MTU per year of high-level wastes disposal in the geologic repository. This document contains both functional descriptions of the processes in the waste management system and physical descriptions of the equipment and facilities necessary for performance of those processes. These descriptions contain the level of detail needed for the projected systems analysis studies. The Repository-Only System contains all system components, from the waste storage facilities of the waste generators to the underground facilities for final disposal of the wastes. The major facilities in the system are the waste generator waste storage facilities, a repository facility that packages the wastes and than emplaces them in the geologic repository, and the transportation equipment and facilities for transporting the wastes between these major facilities. 18 refs., 39 figs

  17. Description, Modelling and Design of Production Systems

    DEFF Research Database (Denmark)

    Jacobsen, Peter; Rudolph, Carsten

    1997-01-01

    Design of production systems are rarely an activity in which decision makers in most production companies have much experience. In future, this activity is to be more recurrent due to more and more frequent changes in the production task. Consequently, the decision makers are in need of better...... management tools and methods for description and modelling of production systems supporting the decisions. In this article a structural framework to describe and model production systems will be introduced, and it is shown how the production system of a minor Danish manufacturer of electromechanical...

  18. On the thermodynamic description of real systems

    International Nuclear Information System (INIS)

    Bernardes, N.

    1984-01-01

    A new method of approach to the theory of the thermodynamic properties of real systems is proposed, to include interactions among the constituent particles of the system. The method consists in obtaining the entropy of a real system from the entropy of the corresponding ideal system by a translation in the internal energy and other relevant extensive variables. The usefulness of the method is displayed by application to the cases of: (i) real gases, and (ii) spin paramagnetism with interactions among spins. It is shown that this description corresponds to a generalization of the molecular field approximation. (Author) [pt

  19. SNF/HLW Transfer System Description Document

    International Nuclear Information System (INIS)

    W. Holt

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF)/high-level radioactive waste (HLW) transfer system and associated bases, which will allow the design effort to proceed to license application. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in this SDD reflects the current results of the design process

  20. System Design Description PFP Thermal Stabilization

    International Nuclear Information System (INIS)

    RISENMAY, H.R.

    2000-01-01

    The purpose of this document is to provide a system design description (SDD) and design basis for the Plutonium Finishing Plant (PFP) Thermal Stabilization project. The chief objective of the SDD is to document the Structures, Systems, and Components (SSCs) that establish and maintain the facility Safety Envelope necessary for normal safe operation of the facility; as identified in the FSAR, the OSRs, and Safety Assessment Documents (SADs). This safety equipment documentation should satisfy guidelines for the SDD given in WHC-SD-CP-TI-18 1, Criteria for Identification and Control of Equipment Necessary for Preservation of the Safety Envelope and Safe Operation of PFP. The basis for operational, alarm response, maintenance, and surveillance procedures are also identified and justified in this document. This document and its appendices address the following elements of the PFP Thermal Stabilization project: Functional and design requirements; Design description; Safety Envelope Analysis; Safety Equipment Class; and Operational, maintenance and surveillance procedures

  1. LANL environmental restoration site ranking system: System description. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Merkhofer, L.; Kann, A.; Voth, M. [Applied Decision Analysis, Inc., Menlo Park, CA (United States)

    1992-10-13

    The basic structure of the LANL Environmental Restoration (ER) Site Ranking System and its use are described in this document. A related document, Instructions for Generating Inputs for the LANL ER Site Ranking System, contains detailed descriptions of the methods by which necessary inputs for the system will be generated. LANL has long recognized the need to provide a consistent basis for comparing the risks and other adverse consequences associated with the various waste problems at the Lab. The LANL ER Site Ranking System is being developed to help address this need. The specific purpose of the system is to help improve, defend, and explain prioritization decisions at the Potential Release Site (PRS) and Operable Unit (OU) level. The precise relationship of the Site Ranking System to the planning and overall budget processes is yet to be determined, as the system is still evolving. Generally speaking, the Site Ranking System will be used as a decision aid. That is, the system will be used to aid in the planning and budgetary decision-making process. It will never be used alone to make decisions. Like all models, the system can provide only a partial and approximate accounting of the factors important to budget and planning decisions. Decision makers at LANL will have to consider factors outside of the formal system when making final choices. Some of these other factors are regulatory requirements, DOE policy, and public concern. The main value of the site ranking system, therefore, is not the precise numbers it generates, but rather the general insights it provides.

  2. LANL environmental restoration site ranking system: System description. Final report

    International Nuclear Information System (INIS)

    Merkhofer, L.; Kann, A.; Voth, M.

    1992-01-01

    The basic structure of the LANL Environmental Restoration (ER) Site Ranking System and its use are described in this document. A related document, Instructions for Generating Inputs for the LANL ER Site Ranking System, contains detailed descriptions of the methods by which necessary inputs for the system will be generated. LANL has long recognized the need to provide a consistent basis for comparing the risks and other adverse consequences associated with the various waste problems at the Lab. The LANL ER Site Ranking System is being developed to help address this need. The specific purpose of the system is to help improve, defend, and explain prioritization decisions at the Potential Release Site (PRS) and Operable Unit (OU) level. The precise relationship of the Site Ranking System to the planning and overall budget processes is yet to be determined, as the system is still evolving. Generally speaking, the Site Ranking System will be used as a decision aid. That is, the system will be used to aid in the planning and budgetary decision-making process. It will never be used alone to make decisions. Like all models, the system can provide only a partial and approximate accounting of the factors important to budget and planning decisions. Decision makers at LANL will have to consider factors outside of the formal system when making final choices. Some of these other factors are regulatory requirements, DOE policy, and public concern. The main value of the site ranking system, therefore, is not the precise numbers it generates, but rather the general insights it provides

  3. Waste receiving and processing plant control system; system design description

    International Nuclear Information System (INIS)

    LANE, M.P.

    1999-01-01

    The Plant Control System (PCS) is a heterogeneous computer system composed of numerous sub-systems. The PCS represents every major computer system that is used to support operation of the Waste Receiving and Processing (WRAP) facility. This document, the System Design Description (PCS SDD), includes several chapters and appendices. Each chapter is devoted to a separate PCS sub-system. Typically, each chapter includes an overview description of the system, a list of associated documents related to operation of that system, and a detailed description of relevant system features. Each appendice provides configuration information for selected PCS sub-systems. The appendices are designed as separate sections to assist in maintaining this document due to frequent changes in system configurations. This document is intended to serve as the primary reference for configuration of PCS computer systems. The use of this document is further described in the WRAP System Configuration Management Plan, WMH-350, Section 4.1

  4. Waste receiving and processing plant control system; system design description

    Energy Technology Data Exchange (ETDEWEB)

    LANE, M.P.

    1999-02-24

    The Plant Control System (PCS) is a heterogeneous computer system composed of numerous sub-systems. The PCS represents every major computer system that is used to support operation of the Waste Receiving and Processing (WRAP) facility. This document, the System Design Description (PCS SDD), includes several chapters and appendices. Each chapter is devoted to a separate PCS sub-system. Typically, each chapter includes an overview description of the system, a list of associated documents related to operation of that system, and a detailed description of relevant system features. Each appendice provides configuration information for selected PCS sub-systems. The appendices are designed as separate sections to assist in maintaining this document due to frequent changes in system configurations. This document is intended to serve as the primary reference for configuration of PCS computer systems. The use of this document is further described in the WRAP System Configuration Management Plan, WMH-350, Section 4.1.

  5. Remote operation system for container

    International Nuclear Information System (INIS)

    Nakahara, Hirotaka; Hayata, Takashi; Kajiyama, Shigeru; Takahashi, Fuminobu

    1998-01-01

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

  6. Safety system for reactor container

    International Nuclear Information System (INIS)

    Shimizu, Miwako; Seki, Osamu; Mano, Takio.

    1995-01-01

    A slanted structure is formed below a reactor core where there is a possibility that molten reactor core materials are dropped, and above a water level of a pool which is formed by coolants flown from a reactor recycling system and accumulated on the inner bottom of the reactor container, to prevent molten fuels from dropping at once in the form of a large amount of lump. The molten materials are provisionally received on the structure, gradually formed into small pieces and then dropped. Further, the molten materials are dropped and received provisionally on a group of coolant-flowing pipelines below the structure, to lower the temperature of the molten materials, and then the reactor core molten materials are gradually formed into small pieces and dropped into the pool water. Since they are not dropped directly into the pool water but dropped gradually into the pool water as small droplets, occurrence of steam explosion can be reduced. The occurrence of steam explosion due to dropped molten reactor core material and pool water is suppressed, and the molten materials are kept in the pool water, thereby enabling to maintain the integrity of the reactor container more effectively. (N.H.)

  7. CDMS: CAD data set system design description. Revision 1

    International Nuclear Information System (INIS)

    Gray, E.L.

    1994-01-01

    This document is intended to formalize the program design of the CAD Data Set Management System (CDMS) and to be the vehicle to communicate the design to the Engineering, Design Services, and Configuration Management organizations and the WHC IRM Analysts/Programmers. The SDD shows how the software system will be structured to satisfy the requirements identified in the WHC-SD-GN-CSRS-30005 CDMS Software Requirement Specification (SRS). It is a description of the software structure, software components, interfaces, and data that make up the CDMS System. The design descriptions contained within this document will describe in detail the software product that will be developed to assist the aforementioned organizations for the express purpose of managing CAD data sets associated with released drawings, replacing the existing locally developed system and laying the foundation for automating the configuration management

  8. Containment and surveillance systems for international safeguards

    International Nuclear Information System (INIS)

    Ney, J.F.

    1978-01-01

    Important criteria in measuring the effectiveness of IAEA safeguards include timeliness of detection of diversion, timeliness of reporting such detections, and confidence in determining the amount of material diverted. Optimum use of IAEA inspectors, combined with adequate instrumentation, can provide a practical means for achieving these criteria. System studies are being carried out for different types of facilities that may come under IAEA safeguards to determine the proper balance between inspector's efforts and the use of safeguards instrumentation. A description of a typical study is presented. Based on the results of these studies, the program undertaken to develop those containment and surveillance subsystems for which the technical feasibility and operational acceptability need to be established is described

  9. Demonstration Advanced Avionics System (DAAS) function description

    Science.gov (United States)

    Bailey, A. J.; Bailey, D. G.; Gaabo, R. J.; Lahn, T. G.; Larson, J. C.; Peterson, E. M.; Schuck, J. W.; Rodgers, D. L.; Wroblewski, K. A.

    1982-01-01

    The Demonstration Advanced Avionics System, DAAS, is an integrated avionics system utilizing microprocessor technologies, data busing, and shared displays for demonstrating the potential of these technologies in improving the safety and utility of general aviation operations in the late 1980's and beyond. Major hardware elements of the DAAS include a functionally distributed microcomputer complex, an integrated data control center, an electronic horizontal situation indicator, and a radio adaptor unit. All processing and display resources are interconnected by an IEEE-488 bus in order to enhance the overall system effectiveness, reliability, modularity and maintainability. A detail description of the DAAS architecture, the DAAS hardware, and the DAAS functions is presented. The system is designed for installation and flight test in a NASA Cessna 402-B aircraft.

  10. Description of the CONTAIN input model for the Dodewaard nuclear power plant

    International Nuclear Information System (INIS)

    Velema, E.J.

    1992-02-01

    This report describes the ECN standard CONTAIN input model for the Dodewaard Nuclear Power Plant (NPP) that has been developed by ECN. This standard input model will serve as a basis for analyses of the phenomena which may occur inside the Dodewaard containment in the event of a postulated severe accident. Boundary conditions for specific containment analyses can easily be implemented in the input model. as a result ECN will be able to respond quickly on requests for analyses from the utilities of the authorities. The report also includes brief descriptions of the Dodewaard NPP and the CONTAIN computer program. (author). 7 refs.; 5 figs.; 3 tabs

  11. System design description PFP thermal stabilization

    International Nuclear Information System (INIS)

    RISENMAY, H.R.

    1998-01-01

    The purpose of this document is to provide a system design description and design basis for the Plutonium Finishing P1ant (PFP) Thermal Stabilization project. The sources of material for this project are residues scraped from glovebox floors and materials already stored in vault storage that need further stabilizing to meet the 3013 storage requirements. Stabilizing this material will promote long term storage and reduced worker exposure. This document addresses: function design, equipment, and safety requirements for thermal stabilization of plutonium residues and oxides

  12. Automatic TLI recognition system. Part 1: System description

    Energy Technology Data Exchange (ETDEWEB)

    Partin, J.K.; Lassahn, G.D.; Davidson, J.R.

    1994-05-01

    This report describes an automatic target recognition system for fast screening of large amounts of multi-sensor image data, based on low-cost parallel processors. This system uses image data fusion and gives uncertainty estimates. It is relatively low cost, compact, and transportable. The software is easily enhanced to expand the system`s capabilities, and the hardware is easily expandable to increase the system`s speed. This volume gives a general description of the ATR system.

  13. Building a secondary containment system

    Energy Technology Data Exchange (ETDEWEB)

    Broder, M.F.

    1994-10-01

    Retail fertilizer and pesticide dealers across the United States are installing secondary containment at their facilities or are seriously considering it. Much of this work is in response to new state regulations; however, many dealers not facing new regulations are upgrading their facilities to reduce their liability, lower their insurance costs, or comply with anticipated regulations. The Tennessee Valley Authority`s (TVA) National Fertilizer and Environmental Research Center (NFERC) has assisted dealers in 22 states in retrofitting containment to their facilities. Simultaneous improvements in the operational efficiency of the facilities have been achieved at many of the sites. This paper is based on experience gained in that work and details the rationale used in planning secondary containment and facility modifications.

  14. Description of an open quantum mechanical system

    International Nuclear Information System (INIS)

    Rotter, I.; Forschungszentrum Rossendorf e.V.

    1994-05-01

    A model for the description of an open quantum mechanical many-particle system is formulated. It starts from the shell model and treats the continuous states by a coupled channels method. The mixing of the discrete shell model states via the continuum of decay channels results in the genuine decaying states of the system. These states are eigenstates of a non-Hermitean Hamilton operator the eigenvalues of which give both the energies and the widths of the states. All correlations between two particles which are caused by the two-particle residual interaction, are taken into account including those via the continuum. In the formalism describing the open quantum mechanical system, the coupling between the system and its environment appears nonlinearly. If the resonance states start to overlap, a redistribution of the spectroscopic values ('trapping effect') takes place. As a result, the complexity of the system is reduced at high level density, structures in space and time are formed. This redistribution describes, on the one hand, the transition from the well-known nuclear properties at low level density to those at high level density and fits, on the other hand, into the concept of selforganization. (orig.)

  15. Containment

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The primary mission of the Containment Group is to ensure that underground nuclear tests are satisfactorily contained. The main goal is the development of sound technical bases for containment-related methodology. Major areas of activity include siting, geologic description, emplacement hole stemming, and phenomenological predictions. Performance results of sanded gypsum concrete plugs on the Jefferson, Panamint, Cornucopia, Labquark, and Bodie events are given. Activities are also described in the following areas: computational capabilities site description, predictive modeling, and cavity-pressure measurement. Containment publications are listed. 8 references

  16. KALIMER fuel system preliminary design description

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, B.O.; Nam, C.; Paek, S.K.

    1998-10-01

    This document provides general design concepts, design basis, preliminary design specification and design technologies which are needed for designing the fuel/non-fuel rods and assembly ducts of the KALIMER fuel system. The core of LMFBR consists of driver fuel assembly, blanket assembly, reflector assembly, shielding assembly, control assembly and GEM (Gas Expansion Module) as well as USS, dummy assembly, detector assembly. These core components must be designed to withstand the high temperature, high flux for a long irradiation exposure time. Due to the high temperature and high flux, irradiation creep and swelling as well as thermal-mechanical deformation are occurred at the fuel/non-fuel system and cause the deformations of materials and the geometric deflections at fuel/non-fuel rods, assembly ducts and components. In order to overcome these intricate phenomena through the engineering design, the design basis including theoretical analysis methodologies and design considerations, material characteristics of fuel system, and the specifications and drawings of fuel/non-fuel rods and assembly ducts, respectively, are presented. This document is preliminary design description which is produced in the conceptual design stage, and does not present the detailed and finalized design data which can be for the manufacturing. (author). 22 refs

  17. System description of the Basic MRS System for the FY 1990 Systems Integration Program studies

    International Nuclear Information System (INIS)

    McKee, R.W.; Young, J.R.; Konzek, G.J.

    1991-07-01

    This document provides both functional and physical descriptions of a conceptual high-level waste management system defined as a Basic MRS System. Its purpose is to provide a basis for required system computer modeling and system studies initiated in FY 1990 under the Systems Integration Program of the Office of Civilian Radioactive Waste Management Office (OCRWM). Two specific systems studies initiated in FY 1990, the Reference System Performance Evaluation and the Aggregate Receipt Rate Study, utilize the information in this document. The Basic MRS System is the current OCRWM reference high-level radioactive wastes repository system concept. It is designed to accept 3000 MTU per year of spent fuel and 400 equivalent MTU per year of high-level wastes. The Basic MRS System includes a storage-only MRS that provides for a limited amount of commercial spent fuel storage capacity prior to acceptance by the geologic repository for disposal. This document contains both functional descriptions of the processes in the waste management system and physical descriptions of the equipment and facilities necessary for performance of those processes. The basic MRS system contains all system components, from the waste storage facilities of the waste generators to the underground facilities for final disposal of the wastes. The major facilities in the system are the waste generator waste storage facilities, an MRS facility that provides interim storage wastes accepted from the waste generators, a repository facility that packages the wastes and then emplaces them in the geologic repository, and the transportation equipment and facilities for transporting the waste between these major facilities

  18. Understanding aging in containment cooling systems

    International Nuclear Information System (INIS)

    Lofaro, R.J.

    1993-01-01

    A study has been performed to assess the effects of aging in nuclear power plant containment cooling systems. Failure records from national databases, as well as plant specific data were reviewed and analyzed to identify aging characteristics for this system. The predominant aging mechanisms were determined, along with the most frequently failed components and their associated failure modes. This paper discusses the aging mechanisms present in the containment spray system and the containment fan cooler system, which are two systems used to provide the containment cooling function. The failure modes, along with the relative frequency of each is also discussed

  19. AHTR Refueling Systems and Process Description

    Energy Technology Data Exchange (ETDEWEB)

    Varma, V.K.; Holcomb, D.E.; Bradley, E.C.; Zaharia, N.M.; Cooper, E.J.

    2012-07-15

    The Advanced High-Temperature Reactor (AHTR) is a design concept for a central station-type [1500 MW(e)] Fluoride salt–cooled High-temperature Reactor (FHR) that is currently undergoing development by Oak Ridge National Laboratory for the US. Department of Energy, Office of Nuclear Energy’s Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR is approaching a preconceptual level of maturity. An initial integrated layout of its major systems, structures, and components (SSCs), and an initial, high-level sequence of operations necessary for constructing and operating the plant is nearing completion. An overview of the current status of the AHTR concept has been recently published [1], and a report providing a more detailed overview of the AHTR structures and mechanical systems is currently in preparation. This report documents the refueling components and processes envisioned at this early development phase. The report is limited to the refueling aspects of the AHTR and does not include overall reactor or power plant design information. The report, however, does include a description of the materials envisioned for the various components and the instrumentation necessary to control the refueling process. The report begins with an overview of the refueling strategy. Next a mechanical description of the AHTR fuel assemblies and core is provided. The reactor vessel upper assemblies are then described. Following this the refueling path structures and the refueling mechanisms and components are described. The sequence of operations necessary to fuel and defuel the reactor is then discussed. The report concludes with a discussion of the

  20. AHTR Refueling Systems and Process Description

    Energy Technology Data Exchange (ETDEWEB)

    Varma, Venugopal Koikal [ORNL; Holcomb, David Eugene [ORNL; Bradley, Eric Craig [ORNL; Zaharia, Nathaniel M [ORNL; Cooper, Eliott J [ORNL

    2012-07-01

    The Advanced High-Temperature Reactor (AHTR) is a design concept for a central station-type [1500 MW(e)] Fluoride salt-cooled High-temperature Reactor (FHR) that is currently undergoing development by Oak Ridge National Laboratory for the US. Department of Energy, Office of Nuclear Energy's Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR is approaching a preconceptual level of maturity. An initial integrated layout of its major systems, structures, and components (SSCs), and an initial, high-level sequence of operations necessary for constructing and operating the plant is nearing completion. An overview of the current status of the AHTR concept has been recently published and a report providing a more detailed overview of the AHTR structures and mechanical systems is currently in preparation. This report documents the refueling components and processes envisioned at this early development phase. The report is limited to the refueling aspects of the AHTR and does not include overall reactor or power plant design information. The report, however, does include a description of the materials envisioned for the various components and the instrumentation necessary to control the refueling process. The report begins with an overview of the refueling strategy. Next a mechanical description of the AHTR fuel assemblies and core is provided. The reactor vessel upper assemblies are then described. Following this the refueling path structures and the refueling mechanisms and components are described. The sequence of operations necessary to fuel and defuel the reactor is then discussed. The report concludes with a discussion of the

  1. Advanced Transport Operating System (ATOPS) color displays software description microprocessor system

    Science.gov (United States)

    Slominski, Christopher J.; Plyler, Valerie E.; Dickson, Richard W.

    1992-01-01

    This document describes the software created for the Sperry Microprocessor Color Display System used for the Advanced Transport Operating Systems (ATOPS) project on the Transport Systems Research Vehicle (TSRV). The software delivery known as the 'baseline display system', is the one described in this document. Throughout this publication, module descriptions are presented in a standardized format which contains module purpose, calling sequence, detailed description, and global references. The global reference section includes procedures and common variables referenced by a particular module. The system described supports the Research Flight Deck (RFD) of the TSRV. The RFD contains eight cathode ray tubes (CRTs) which depict a Primary Flight Display, Navigation Display, System Warning Display, Takeoff Performance Monitoring System Display, and Engine Display.

  2. FFTF-containment air-cleaning system

    International Nuclear Information System (INIS)

    Mahaffey, M.K.; Stepnewski, D.D.

    1981-01-01

    The FFTF Containment can accommodate all design basis events and the hypothetical core disruptive accident with adequate margin without a venting or purging system; however, in concert with the development objective, a system was designed and constructed to evaluate technology related to containment atmosphere venting and cleanup functions. The system can be used to purge high H 2 concentrations or to vent excessive containment pressure. In either case containment atmosphere is exhausted through an aqueous scrubber system consisting of a venturi scrubber and fibrous filter bank

  3. Remote container monitoring and surveillance systems

    International Nuclear Information System (INIS)

    Resnik, W.M.; Kadner, S.P.

    1995-01-01

    Aquila Technologies Group is developing a monitoring and surveillance system to monitor containers of nuclear materials. The system will both visually and physically monitor the containers. The system is based on the combination of Aquila's Gemini All-Digital Surveillance System and on Aquila's AssetLAN trademark asset tracking technology. This paper discusses the Gemini Digital Surveillance system as well as AssetLAN technology. The Gemini architecture with emphasis on anti-tamper security features is also described. The importance of all-digital surveillance versus other surveillance methods is also discussed. AssetLAN trademark technology is described, emphasizing the ability to continually track containers (as assets) by location utilizing touch memory technology. Touch memory technology provides unique container identification, as well as the ability to store and retrieve digital information on the container. This information may relate to container maintenance, inspection schedules, and other information. Finally, this paper describes the combination of the Gemini system with AssetLAN technology, yielding a self contained, container monitoring and area/container surveillance system. Secure container fixture design considerations are discussed. Basic surveillance review functions are also discussed

  4. Advanced Transport Operating System (ATOPS) color displays software description: MicroVAX system

    Science.gov (United States)

    Slominski, Christopher J.; Plyler, Valerie E.; Dickson, Richard W.

    1992-01-01

    This document describes the software created for the Display MicroVAX computer used for the Advanced Transport Operating Systems (ATOPS) project on the Transport Systems Research Vehicle (TSRV). The software delivery of February 27, 1991, known as the 'baseline display system', is the one described in this document. Throughout this publication, module descriptions are presented in a standardized format which contains module purpose, calling sequence, detailed description, and global references. The global references section includes subroutines, functions, and common variables referenced by a particular module. The system described supports the Research Flight Deck (RFD) of the TSRV. The RFD contains eight Cathode Ray Tubes (CRTs) which depict a Primary Flight Display, Navigation Display, System Warning Display, Takeoff Performance Monitoring System Display, and Engine Display.

  5. Igloo containment system for improvised explosive devices

    International Nuclear Information System (INIS)

    Dyckes, G.W.

    1980-09-01

    A method for containing or partially containing the blast and dispersal of radioactive particulate from improvised explosive devices is described. The containment system is restricted to devices located in fairly open areas at ground level, e.g., devices concealed in trucks, vans, transportainers, or small buildings which are accessible from all sides

  6. Integrated Visualisation and Description of Complex Systems

    National Research Council Canada - National Science Library

    Goodburn, D

    1999-01-01

    .... Guided by a conceptual model of a description process that is driven by user information needs within a domain context, the approach incorporates the use of novel visualization techniques based...

  7. NucleDyne's passive containment system

    International Nuclear Information System (INIS)

    Falls, O.B. Jr.; Kleimola, F.W.

    1987-01-01

    A simple definition of the passive containment system is that it is a total safeguards system for light water reactors designed to prevent and contain any accidental release of radioactivity. Its passive features utilize the natural laws of physics and thermodynamics. The system encompasses three basic containments constructed as one integrated structure on the reactor building foundation. The primary containment encloses the reactor pressure vessel and coolant system and passive engineered safety systems and components. Auxiliary containment enclosures house auxiliary systems and components. Secondary containment (the reactor building), housing the primary and auxiliary containment structures, provides a second containment barrier as added defense-in-depth against leakage of radioactivity for all accidents assumed by the industry. The generic features of the passive containment system are applicable to both the boiling water reactors and the pressurized water reactors as standardized features for all power ranges. These features provide for a zero source term, the industry's ultimate safety goal. This paper relates to a four-loop pressurized water reactor

  8. Containments for consolidated nuclear steam systems

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    A containment system for a consolidated nuclear steam system incorporating a nuclear core, steam generator and reactor coolant pumps within a single pressure vessel is described which is designed to provide radiation shielding and pressure suppression. Design details, including those for the dry well and wet well of the containment, are given. (UK)

  9. CONTEMPT4/MOD2: a multicompartment containment system analysis program

    International Nuclear Information System (INIS)

    Metcalfe, L.J.; Mings, W.J.; Hartman, J.E.; Crail, A.C.

    1978-02-01

    CONTEMPT4/MOD2 is a digital computer program, written in FORTRAN IV, which describes the behavior of multicompartment pressurized water reactor (PWR) containment systems and experimental containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and intercompartment mass and energy exchange based on user-supplied values for compartment descriptions, time step and edit controls, and selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, fan coolers, heat conducting structures, sump drain, and PWR ice condensers. Dynamic storage allocations (DSA) is used to limit the amount of computer core used for each problem. Optional automatic time step control allows the code to determine time step sizes within limits dictated by the user. Multicompartment capability (up to 999 individual compartments) and generalized, user-oriented input data descriptions permit improved flexibility over previous codes in the CONTEMPT series. Analytical model descriptions, input instructions, and sample problem results are presented

  10. CONTEMPT 4/MOD 3: a multicompartment containment system analysis program

    International Nuclear Information System (INIS)

    Cheng, T.C.; Metcalfe, L.J.; Hartman, J.E.; Mings, W.J.; Crail, A.C.

    1982-12-01

    CONTEMPT4/MOD3 is a digital computer program, written in FORTRAN IV, that describes the behavior of multicompartment pressurized water reactor (PWR) containment systems and experimental containment systems subjected to postulated loss-of-coolant accident (LOCA) conditons. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and intercompartment mass and energy exchange based on user-supplied values for compartment descriptions, time step and edit controls, and selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, fan coolers, heat-conducting structures, sump drains, and PWR ice condensers. Dynamic stroage allocation (DSA) is used to limit the amount of computer core used for each problem. Optional automatic time step control allows the code to determine time step sizes within limits dictated by the user. Multicompartment capability (up to 999 individual compartments) and generalized, user-oriented input-data descriptions permit improved flexibility over previous codes in the CONTEMPT series. Analytical model descriptions, input instructions, and sample problem results are presented

  11. Description of surface systems. Preliminary site description Simpevarp sub area - Version 1.2

    Energy Technology Data Exchange (ETDEWEB)

    Lindborg, Tobias [ed.

    2005-03-01

    Swedish Nuclear Fuel and Waste Management Co is currently conducting site characterisation in the Simpevarp area. The area is divided into two subareas, the Simpevarp and the Laxemar subarea. The two subareas are surrounded by a common regional model area, the Simpevarp area. This report describes both the regional area and the subareas. This report is an interim version (model version 1.2) of the description of the surface systems at the Simpevarp area, and should be seen as a background report to the site description of the Simpevarp area, version 1.2, SKB-R--05-08. The basis for this description is quality-assured field data available in the SKB SICADA and GIS databases, together with generic data from the literature. The Surface system, here defined as everything above the bedrock, comprises a number of separate disciplines (e.g. hydrology, geology, topography, oceanography and ecology). Each discipline has developed descriptions and models for a number of properties that together represent the site description. The current methodology for developing the surface system description and the integration to ecosystem models is documented in a methodology strategy report SKB-R--03-06. The procedures and guidelines given in that report were followed in this report. Compared with version 1.1 of the surface system description SKB-R--04-25, this report presents considerable additional features, especially in the ecosystem description (Chapter 4) and in the description of the surface hydrology (Section 3.4). A first attempt has also been made to connect the flow of matter (carbon) between the different ecosystems into an overall ecosystem model at a landscape level. A summarised version of this report is also presented in SKB-R--05-08 together with geological-, hydrogeological-, transport properties-, thermal properties-, rock mechanics- and hydrogeochemical descriptions.

  12. Description of surface systems. Preliminary site description Simpevarp sub area - Version 1.2

    International Nuclear Information System (INIS)

    Lindborg, Tobias

    2005-03-01

    Swedish Nuclear Fuel and Waste Management Co is currently conducting site characterisation in the Simpevarp area. The area is divided into two subareas, the Simpevarp and the Laxemar subarea. The two subareas are surrounded by a common regional model area, the Simpevarp area. This report describes both the regional area and the subareas. This report is an interim version (model version 1.2) of the description of the surface systems at the Simpevarp area, and should be seen as a background report to the site description of the Simpevarp area, version 1.2, SKB-R--05-08. The basis for this description is quality-assured field data available in the SKB SICADA and GIS databases, together with generic data from the literature. The Surface system, here defined as everything above the bedrock, comprises a number of separate disciplines (e.g. hydrology, geology, topography, oceanography and ecology). Each discipline has developed descriptions and models for a number of properties that together represent the site description. The current methodology for developing the surface system description and the integration to ecosystem models is documented in a methodology strategy report SKB-R--03-06. The procedures and guidelines given in that report were followed in this report. Compared with version 1.1 of the surface system description SKB-R--04-25, this report presents considerable additional features, especially in the ecosystem description (Chapter 4) and in the description of the surface hydrology (Section 3.4). A first attempt has also been made to connect the flow of matter (carbon) between the different ecosystems into an overall ecosystem model at a landscape level. A summarised version of this report is also presented in SKB-R--05-08 together with geological-, hydrogeological-, transport properties-, thermal properties-, rock mechanics- and hydrogeochemical descriptions

  13. Performance of Sequoyah Containment Anchorage System

    International Nuclear Information System (INIS)

    Fanous, F.; Greimann, L.; Wassef, W.; Bluhm, D.

    1993-01-01

    Deformation of a steel containment anchorage system during a severe accident may result in a leakage path at the containment boundaries. Current design criteria are based on either ductile or brittle failure modes of headed bolts that do not account for factors such as cracking of the containment basemat or deformation of the anchor bolt that may affect the behavior of the containment anchorage system. The purpose of this study was to investigate the performance of a typical ice condenser containment's anchorage system. This was accomplished by analyzing the Sequoyah Containment Anchorage System. Based on a strength of materials approach and assuming that the anchor bolts are resisting the uplift caused by the internal pressure, one can estimate that the failure of the anchor bolts would occur at a containment pressure of 79 psig. To verify these results and to calibrate the strength of materials equation, the Sequoyah containment anchorage system was analyzed with the ABAQUS program using a three-dimensional, finite-element model. The model included portions of the steel containment building, shield building, anchor bolt assembly, reinforced concrete mat and soil foundation material

  14. Reliability analysis of containment isolation systems

    International Nuclear Information System (INIS)

    Pelto, P.J.; Counts, C.A.

    1984-06-01

    The Pacific Northwest Laboratory (PNL) is reviewing available information on containment systems design, operating experience, and related research as part of a project being conducted by the Division of Systems Integration, US Nuclear Regulatory Commission. The basic objective of this work is to collect and consolidate data relevant to assessing the functional performance of containment isolation systems and to use this data to the extent possible to characterize containment isolation system reliability for selected reference designs. This paper summarizes the results from initial efforts which focused on collection of data from available documents and briefly describes detailed review and analysis efforts which commenced recently. 5 references

  15. The Soviet RBMK-1000 containment system

    International Nuclear Information System (INIS)

    Joosten, J.K.

    1988-01-01

    Following the accident in April, 1986, considerable attention was focused on the failure of the containment at the Chernobyl RBMK-1000 nuclear power plant. Conflicting statements arose regarding the nature of the plant's containment system primarily because of terminology differences, translation difficulties and lack of reliable information. This article, based on reports and briefings by the Soviet delegation, during the post-accident review meetings in Vienna and prior publications is intended to clarify perceptions of the Soviet RMBK-1000 nuclear power plant containment system design, and its relevance to containment management concepts. (author)

  16. Cold Vacuum Drying Instrument Air System Design Description. System 12

    International Nuclear Information System (INIS)

    SHAPLEY, B.J.; TRAN, Y.S.

    2000-01-01

    This system design description (SDD) addresses the instrument air (IA) system of the spent nuclear fuel (SNF). This IA system provides instrument quality air to the Cold Vacuum Drying (CVD) Facility. The IA system is a general service system that supports the operation of the heating, ventilation, and air conditioning (HVAC) system, the process equipment skids, and process instruments in the CVD Facility. The following discussion is limited to the compressor, dryer, piping, and valving that provide the IA as shown in Drawings H-1-82222, Cold Vacuum Drying Facility Mechanical Utilities Compressed and Instrument Air PandID, and H-1.82161, Cold Vacuum Drying Facility Process Equipment Skid PandID MCO/Cusk Interface. Figure 1-1 shows the physical location of the 1A system in the CVD Facility

  17. System design description for Waste Information and Control System

    International Nuclear Information System (INIS)

    Harris, R.R.

    1994-01-01

    The Westinghouse Hanford Company (WHC) Hazardous Material Control Group (HMC) of the 222-S Laboratory has requested the development of a system to help resolve many of the difficulties associated with tracking and data collection of containers and drums of waste. This system has been identified as the Waste Information and Control System (WICS). WICS shall partially automate the procedure for acquisition, tracking and reporting of the container, drum, and waste data that is currently manually processed. The WICS project shall use handheld computer units (HCU) to collect laboratory data, a local database with an user friendly interface to import the laboratory data from the HCUs, and barcode technology with associated software and operational procedures. After the container, drum, and waste data has been collected and verified, WICS shall be manipulated to provide informal reports containing data required to properly document waste disposal. 8 refs, 82 figs, 69 tabs

  18. System for inspection of stacked cargo containers

    Science.gov (United States)

    Derenzo, Stephen [Pinole, CA

    2011-08-16

    The present invention relates to a system for inspection of stacked cargo containers. One embodiment of the invention generally comprises a plurality of stacked cargo containers arranged in rows or tiers, each container having a top, a bottom a first side, a second side, a front end, and a back end; a plurality of spacers arranged in rows or tiers; one or more mobile inspection devices for inspecting the cargo containers, wherein the one or more inspection devices are removeably disposed within the spacers, the inspection means configured to move through the spacers to detect radiation within the containers. The invented system can also be configured to inspect the cargo containers for a variety of other potentially hazardous materials including but not limited to explosive and chemical threats.

  19. System design description for the whole element furnace testing system

    International Nuclear Information System (INIS)

    Ritter, G.A.; Marschman, S.C.; MacFarlan, P.J.; King, D.A.

    1998-05-01

    This document provides a detailed description of the Hanford Spent Nuclear Fuel (SNF) Whole Element Furnace Testing System located in the Postirradiation Testing Laboratory G-Cell (327 Building). Equipment specifications, system schematics, general operating modes, maintenance and calibration requirements, and other supporting information are provided in this document. This system was developed for performing cold vacuum drying and hot vacuum drying testing of whole N-Reactor fuel elements, which were sampled from the 105-K East and K West Basins. The proposed drying processes are intended to allow dry storage of the SNF for long periods of time. The furnace testing system is used to evaluate these processes by simulating drying sequences with a single fuel element and measuring key system parameters such as internal pressures, temperatures, moisture levels, and off-gas composition

  20. Loop containment (joint integrity) assessment Brayton Isotope Power System flight system

    International Nuclear Information System (INIS)

    1976-01-01

    The Brayton Isotope Power System (BIPS) contains a large number of joints. Since the failure of a joint would result in loss of the working fluid and consequential failure of the BIPS, the integrity of the joints is of paramount importance. The reliability of the ERDA BIPS loop containment (joint integrity) is evaluated. The conceptual flight system as presently configured is depicted. A brief description of the flight system is given

  1. Self-contained microfluidic systems: a review.

    Science.gov (United States)

    Boyd-Moss, Mitchell; Baratchi, Sara; Di Venere, Martina; Khoshmanesh, Khashayar

    2016-08-16

    Microfluidic systems enable rapid diagnosis, screening and monitoring of diseases and health conditions using small amounts of biological samples and reagents. Despite these remarkable features, conventional microfluidic systems rely on bulky expensive external equipment, which hinders their utility as powerful analysis tools outside of research laboratories. 'Self-contained' microfluidic systems, which contain all necessary components to facilitate a complete assay, have been developed to address this limitation. In this review, we provide an in-depth overview of self-contained microfluidic systems. We categorise these systems based on their operating mechanisms into three major groups: passive, hand-powered and active. Several examples are provided to discuss the structure, capabilities and shortcomings of each group. In particular, we discuss the self-contained microfluidic systems enabled by active mechanisms, due to their unique capability for running multi-step and highly controllable diagnostic assays. Integration of self-contained microfluidic systems with the image acquisition and processing capabilities of smartphones, especially those equipped with accessory optical components, enables highly sensitive and quantitative assays, which are discussed. Finally, the future trends and possible solutions to expand the versatility of self-contained, stand-alone microfluidic platforms are outlined.

  2. Integrated Visualisation and Description of Complex Systems

    National Research Council Canada - National Science Library

    Goodburn, D

    1999-01-01

    ... on system topographies and feature overlays. System information from the domain's information space is filtered and integrated into a Composite Systems Model that provides a basis for consistency and integration between all system views...

  3. Passive containment system for a nuclear reactor

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1976-01-01

    A containment system is described that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is continuously maintained submerged in liquid. The primary containment vessel is restored to a high subatmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means

  4. Simplified safety and containment systems for the iris reactor

    International Nuclear Information System (INIS)

    Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

  5. Sewage Solids Irradiator Transportation System (SSITS) cask: preliminary design description

    International Nuclear Information System (INIS)

    Eakes, R.G.; Kempka, S.N.; Lamoreaux, G.H.; Sutherland, S.H.

    1983-02-01

    The preliminary design of the Sewage Solids Irradiator Transportation System (SSITS) Cask is presented in this document. The SSITS cask is to be used for the transport of radioactive cesium chloride and strontium fluoride capsules which are of use in irradiators or as heat sources. The SSITS cask is approximately 1.4 m in diameter, 1.3 m high, weighs roughly 9 t, provides 33 cm of steel shielding, and can dissipate up to 5.2 kW of decay heat. The cask design criteria are identified and a description of the cask design and operation is provided. Detailed analyses of the design were performed to demonstrate licensability of the cask by the Nuclear Regulatory Commission (NRC). Results of the analyses indicate that the preliminary design is in compliance with the pertinent regulatory requirements for licensing of a radioactive material transportation container

  6. Cold Vacuum Drying Facility Crane and Hoist System Design Description. System 14

    International Nuclear Information System (INIS)

    TRAN, Y.S.

    2000-01-01

    This system design description (SDD) is for the Cold Vacuum Drying (CVD) Facility overhead crane and hoist system. The overhead crane and hoist system is a general service system. It is located in the process bays of the CVD Facility, supports the processes required to drain the water and dry the spent nuclear fuel (SNF) contained in the multi-canister overpacks (MCOs) after they have been removed from the K-Basins. The location of the system in the process bay is shown

  7. CONTEMPT4/MOD6: a multicompartment containment system analysis program

    International Nuclear Information System (INIS)

    Lin, C.C.; Economos, C.; Lehner, J.R.; Maise, G.

    1986-03-01

    CONTEMPT4/MOD6 is a digital computer program that describes the response of multicompartment containment system subjected to postulated loss-of-coolant accident (LOCA) conditions. The program is written in FORTRAN IV and can accomodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures and mass and energy inventories due to intercompartment mass and energy exchange taking into account user supplied descriptions of compartments, intercompartment junction flow areas, LOCA source terms and user selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers and BWR pressure suppression systems. To accommodate degraded core type accidents, analytical models for hydrogen and carbon monoxide combustion within compartments and energy transfer due to gas radiation are also provided. Dynamic storage allocation (DSA) is used to limit the amount of computer core used for each problem. The flexibility needed to more realistically model the complexity of prototypical containments is provided by the multicompartment capability (up to 999 individual compartments) and generalized user oriented input data descriptions. The program employs an implicit algorithm to compute junction flow when numerically induced flow oscillations are encountered. This capability provides significant reduction of computer run time relative to previous codes in the CONTEMPT series. Descriptions of these analytical models are presented, together with input instructions for the CONTEMPT4/MOD6 program and sample problem results. 23 refs., 62 figs

  8. Integrated environment, safety, and health management system description

    International Nuclear Information System (INIS)

    Zoghbi, J. G.

    2000-01-01

    The Integrated Environment, Safety, and Health Management System Description that is presented in this document describes the approach and management systems used to address integrated safety management within the Richland Environmental Restoration Project

  9. WASTE TREATMENT BUILDING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    F. Habashi

    2000-06-22

    The Waste Treatment Building System provides the space, layout, structures, and embedded subsystems that support the processing of low-level liquid and solid radioactive waste generated within the Monitored Geologic Repository (MGR). The activities conducted in the Waste Treatment Building include sorting, volume reduction, and packaging of dry waste, and collecting, processing, solidification, and packaging of liquid waste. The Waste Treatment Building System is located on the surface within the protected area of the MGR. The Waste Treatment Building System helps maintain a suitable environment for the waste processing and protects the systems within the Waste Treatment Building (WTB) from most of the natural and induced environments. The WTB also confines contaminants and provides radiological protection to personnel. In addition to the waste processing operations, the Waste Treatment Building System provides space and layout for staging of packaged waste for shipment, industrial and radiological safety systems, control and monitoring of operations, safeguards and security systems, and fire protection, ventilation and utilities systems. The Waste Treatment Building System also provides the required space and layout for maintenance activities, tool storage, and administrative facilities. The Waste Treatment Building System integrates waste processing systems within its protective structure to support the throughput rates established for the MGR. The Waste Treatment Building System also provides shielding, layout, and other design features to help limit personnel radiation exposures to levels which are as low as is reasonably achievable (ALARA). The Waste Treatment Building System interfaces with the Site Generated Radiological Waste Handling System, and with other MGR systems that support the waste processing operations. The Waste Treatment Building System interfaces with the General Site Transportation System, Site Communications System, Site Water System, MGR

  10. WASTE TREATMENT BUILDING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    Habashi, F.

    2000-01-01

    The Waste Treatment Building System provides the space, layout, structures, and embedded subsystems that support the processing of low-level liquid and solid radioactive waste generated within the Monitored Geologic Repository (MGR). The activities conducted in the Waste Treatment Building include sorting, volume reduction, and packaging of dry waste, and collecting, processing, solidification, and packaging of liquid waste. The Waste Treatment Building System is located on the surface within the protected area of the MGR. The Waste Treatment Building System helps maintain a suitable environment for the waste processing and protects the systems within the Waste Treatment Building (WTB) from most of the natural and induced environments. The WTB also confines contaminants and provides radiological protection to personnel. In addition to the waste processing operations, the Waste Treatment Building System provides space and layout for staging of packaged waste for shipment, industrial and radiological safety systems, control and monitoring of operations, safeguards and security systems, and fire protection, ventilation and utilities systems. The Waste Treatment Building System also provides the required space and layout for maintenance activities, tool storage, and administrative facilities. The Waste Treatment Building System integrates waste processing systems within its protective structure to support the throughput rates established for the MGR. The Waste Treatment Building System also provides shielding, layout, and other design features to help limit personnel radiation exposures to levels which are as low as is reasonably achievable (ALARA). The Waste Treatment Building System interfaces with the Site Generated Radiological Waste Handling System, and with other MGR systems that support the waste processing operations. The Waste Treatment Building System interfaces with the General Site Transportation System, Site Communications System, Site Water System, MGR

  11. Description of surface systems. Preliminary site description. Forsmark area Version 1.2

    Energy Technology Data Exchange (ETDEWEB)

    Lindborg, Tobias [ed.

    2005-06-01

    Swedish Nuclear Fuel and Waste Management Co (SKB) started site investigations for a deep repository for spent nuclear fuel in 2002 at two different sites in Sweden, Forsmark and Oskarshamn. The investigations should provide necessary information for a license application aimed at starting underground exploration. For this reason, ecosystem data need to be interpreted and assessed into site descriptive models, which in turn are used for safety assessment studies and for environmental impact assessment. Descriptions of the surface system are also needed for further planning of the site investigations. This report describes the surface ecosystems of the Forsmark site (e.g. hydrology, Quaternary deposits, chemistry, vegetation, animals and the human land use). The ecosystem description is an integration of the site and its regional setting, covering the current state of the biosphere as well as the ongoing natural processes affecting the longterm development. Improving the descriptions is important during both the initial and the complete site investigation phase. Before starting of the initial phase in Forsmark, version 0 of the site descriptive model was developed. The results of the initial site investigation phase is compiled into a preliminary site description of Forsmark (version 1.2) in June 2005. This report provides the major input and background to the biosphere description, in the 1.2 version of the Forsmark site description. The basis for this interim version is quality-assured field data from the Forsmark sub area and regional area, available in the SKB SICADA, and GIS data bases as of July 31th 2004 as well as version 1.1 of the Site Descriptive Model. To achieve an ecosystem site description there is a need to develop discipline-specific models by interpreting and analysing primary data. The different discipline-specific models are then integrated into a system describing interactions and flows and stocks of matter between and within functional units in

  12. Description of surface systems. Preliminary site description. Forsmark area Version 1.2

    International Nuclear Information System (INIS)

    Lindborg, Tobias

    2005-06-01

    Swedish Nuclear Fuel and Waste Management Co (SKB) started site investigations for a deep repository for spent nuclear fuel in 2002 at two different sites in Sweden, Forsmark and Oskarshamn. The investigations should provide necessary information for a license application aimed at starting underground exploration. For this reason, ecosystem data need to be interpreted and assessed into site descriptive models, which in turn are used for safety assessment studies and for environmental impact assessment. Descriptions of the surface system are also needed for further planning of the site investigations. This report describes the surface ecosystems of the Forsmark site (e.g. hydrology, Quaternary deposits, chemistry, vegetation, animals and the human land use). The ecosystem description is an integration of the site and its regional setting, covering the current state of the biosphere as well as the ongoing natural processes affecting the longterm development. Improving the descriptions is important during both the initial and the complete site investigation phase. Before starting of the initial phase in Forsmark, version 0 of the site descriptive model was developed. The results of the initial site investigation phase is compiled into a preliminary site description of Forsmark (version 1.2) in June 2005. This report provides the major input and background to the biosphere description, in the 1.2 version of the Forsmark site description. The basis for this interim version is quality-assured field data from the Forsmark sub area and regional area, available in the SKB SICADA, and GIS data bases as of July 31th 2004 as well as version 1.1 of the Site Descriptive Model. To achieve an ecosystem site description there is a need to develop discipline-specific models by interpreting and analysing primary data. The different discipline-specific models are then integrated into a system describing interactions and flows and stocks of matter between and within functional units in

  13. JERICHO computer code: PWR containment response during severe accidents description and sensitivity analysis

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.

    1983-12-01

    The JERICHO code has been developed in order to study the thermodynamic behaviour inside the reactor containment building for the complete spectrum of accident sequences likely to occur in such a reactor, including models for the various mass and energy transfer phenomena, for water spray, for hydrogen and carbon monoxide flammability limits and combustion, as well as for containment venting. Sensitivity analyses have been performed on a severe accident sequence, (namely, small LOCA with failure of the emergency core cooling and containment spray systems), involving core melting and subsequent concrete containment basemat erosion. The effect of various models, such as mass and energy transfer to the structures, has been studied. The influence of the concrete composition, of the fission product deposition and of the thermal degradation of the reactor cavity concrete walls on long term thermodynamic behaviour has also been investigated

  14. Passive containment system in high earthquake motion

    International Nuclear Information System (INIS)

    Kleimola, F.W.; Falls, O.B. Jr.

    1977-01-01

    High earthquake motion necessitates major design modifications in the complex of plant structures, systems and components in a nuclear power plant. Distinctive features imposed by seismic category, safety class and quality classification requirements for the high seismic ground acceleration loadings significantly reflect in plant costs. The design features in the Passive Containment System (PCS) responding to high earthquake ground motion are described

  15. Commissioning Ventilated Containment Systems in the Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    2008-08-01

    This Best Practices Guide focuses on the specialized approaches required for ventilated containment systems, understood to be all components that drive and control ventilated enclosures and local exhaust systems within the laboratory. Geared toward architects, engineers, and facility managers, this guide provides information about technologies and practices to use in designing, constructing, and operating operating safe, sustainable, high-performance laboratories.

  16. Active containment systems incorporating modified pillared clays

    International Nuclear Information System (INIS)

    Lundie, P.; McLeod, N.

    1997-01-01

    The application of treatment technologies in active containment systems provides a more advanced and effective method for the remediation of contaminated sites. These treatment technologies can be applied in permeable reactive walls and/or funnel and gate systems. The application of modified pillared clays in active containment systems provides a mechanism for producing permeable reactive walls with versatile properties. These pillared clays are suitably modified to incorporate reactive intercalatants capable of reacting with both a broad range of organic pollutants of varying molecular size, polarity and reactivity. Heavy metals can be removed from contaminated water by conventional ion-exchange and other reactive processes within the clay structure. Complex contamination problems can be addressed by the application of more than one modified clay on a site specific basis. This paper briefly describes the active containment system and the structure/chemistry of the modified pillared clay technology, illustrating potential applications of the in-situ treatment process for contaminated site remediation

  17. Reliability analysis of containment isolation systems

    International Nuclear Information System (INIS)

    Pelto, P.J.; Ames, K.R.; Gallucci, R.H.

    1985-06-01

    This report summarizes the results of the Reliability Analysis of Containment Isolation System Project. Work was performed in five basic areas: design review, operating experience review, related research review, generic analysis and plant specific analysis. Licensee Event Reports (LERs) and Integrated Leak Rate Test (ILRT) reports provided the major sources of containment performance information used in this study. Data extracted from LERs were assembled into a computer data base. Qualitative and quantitative information developed for containment performance under normal operating conditions and design basis accidents indicate that there is room for improvement. A rough estimate of overall containment unavailability for relatively small leaks which violate plant technical specifications is 0.3. An estimate of containment unavailability due to large leakage events is in the range of 0.001 to 0.01. These estimates are dependent on several assumptions (particularly on event duration times) which are documented in the report

  18. Descriptions of Transit Maintenance Management Information Systems

    Science.gov (United States)

    1984-10-01

    This 289-page report presents an overview of ten maintenance management information systems that are currently operational or near-operational on minicomputers or microcomputers. The systems described address one or more of the following concerns: co...

  19. Nova pulse power system description and status

    International Nuclear Information System (INIS)

    Holloway, R.W.; Whitham, K.; Merritt, B.T.; Gritton, D.G.; Oicles, J.A.

    1981-01-01

    The Nova laser system is designed to produce critical data in the nation's inertial confinement fusion effort. It is the world's largest peak power laser and presents various unique pulse power problems. In this paper, pulse power systems for this laser are described, the evolutionary points from prior systems are pointed out, and the current status of the hardware is given

  20. DESCRIPTION OF THE RHIC SEQUENCER SYSTEM

    International Nuclear Information System (INIS)

    DOTTAVIO, T.; FRAK, B.; MORRIS, J.; SATOGATA, T.; VAN ZEIJTS, J.

    2001-01-01

    The movement of the Relativistic Heavy Ion Collider (RHIC) through its various states (eg. injection, acceleration, storage, collisions) is controlled by an application called the Sequencer. This program orchestrates most magnet and instrumentation systems and is responsible for the coordinated acquisition and saving of data from various systems. The Sequencer system, its software infrastructure, support programs, and the language used to drive it are discussed in this paper. Initial operational experience is also described

  1. Description of the grout system dynamic simulation

    International Nuclear Information System (INIS)

    Zimmerman, B.D.

    1993-07-01

    The grout system dynamic computer simulation was created to allow investigation of the ability of the grouting system to meet established milestones, for various assumed system configurations and parameters. The simulation simulates the movement of tank waste through the system versus time, from initial storage tanks, through feed tanks and the grout plant, then finally to a grout vault. The simulation properly accounts for the following (1) time required to perform various actions or processes, (2) delays involved in gaining regulatory approval, (3) random system component failures, (4) limitations on equipment capacities, (5) available parallel components, and (6) different possible strategies for vault filling. The user is allowed to set a variety of system parameters for each simulation run. Currently, the output of a run primarily consists of a plot of projected grouting campaigns completed versus time, for comparison with milestones. Other outputs involving any model component can also be quickly created or deleted as desired. In particular, sensitivity runs where the effect of varying a model parameter (flow rates, delay times, number of feed tanks available, etc.) on the ability of the system to meet milestones can be made easily. The grout system simulation was implemented using the ITHINK* simulation language for Macintosh** computers

  2. CAMAC instrumentation system: introduction and general description

    International Nuclear Information System (INIS)

    Costrell, L.

    1976-01-01

    The CAMAC instrumentation system is described in a general way in this introductory paper which is followed by papers that discuss the system in greater detail. This paper is an updated version of the introductory paper that appeared in the April 1973 IEEE Transactions on Nuclear Science

  3. Evaluation of the nucledyne passive containment system

    International Nuclear Information System (INIS)

    1981-04-01

    This reports contains: (1) an evaluation by Gilbert/Commonwealth (G/C) of the NucleDyne passive Containment System (PCS) as that conceptual design is applied to a Westinghouse, two loop, Pressurized Water Reactor; (2) an evaluation by Westinghouse of two questions about the impact of the PCS on the Nuclear Steam Supply System (NSSS), which were posed by G/C and best answered by an NSSS vendor; and (3) replies to both the Gilbert/Commonwealth report and the Westinghoue report by NucleDyne Engineering Corporation

  4. Non-perturbative description of quantum systems

    CERN Document Server

    Feranchuk, Ilya; Le, Van-Hoang; Ulyanenkov, Alexander

    2015-01-01

    This book introduces systematically the operator method for the solution of the Schrödinger equation. This method permits to describe the states of quantum systems in the entire range of parameters of Hamiltonian with a predefined accuracy. The operator method is unique compared with other non-perturbative methods due to its ability to deliver in zeroth approximation the uniformly suitable estimate for both ground and excited states of quantum system. The method has been generalized for the application to quantum statistics and quantum field theory.  In this book, the numerous applications of operator method for various physical systems are demonstrated. Simple models are used to illustrate the basic principles of the method which are further used for the solution of complex problems of quantum theory for many-particle systems. The results obtained are supplemented by numerical calculations, presented as tables and figures.

  5. Description of the Energy System of Spain

    Energy Technology Data Exchange (ETDEWEB)

    Caldes, N; Lechon, Y; Labriet, M; Cabal, H; Rua, C de la; Saez, R; Varela, M

    2008-07-01

    The objective of this report is to describe the complete Spain energy system, in order to make possible its modelling with the TIMES model within the NEEDS project (http://www.needs-project.org). (Author) 56 refs.

  6. System Description: Embedding Verification into Microsoft Excel

    OpenAIRE

    Collins, Graham; Dennis, Louise Abigail

    2000-01-01

    The aim of the PROSPER project is to allow the embedding of existing verification technology into applications in such a way that the theorem proving is hidden, or presented to the end user in a natural way. This paper describes a system built to test whether the PROSPER toolkit satisfied this aim. The system combines the toolkit with Microsoft Excel, a popular commercial spreadsheet application.

  7. Anti-foam System design description

    International Nuclear Information System (INIS)

    White, M.A.

    1994-01-01

    The Anti-foam System is a sub-system of the 242-A Evaporator facility. The Anti-foam is used within the C-A-1 Vapor-Liquid Separator, to reduce the effect of foaming and reduce fluid bumping while the vapor and liquid are separated within the C-A-1 Vapor-Liquid Separator. Excessive foaming within the vessel may possibly cause the liquid slurry mixture in the evaporator vessel to foul the de-entrainment pads and cause plant shutdown. The Anti-foam System consists of the following primary elements: the Anti-foam Tank and the Metering Pump. The upgrades to Anti-foam System include the following: installation of a new pump, instruments, and valves; and connection of the instruments, pump and agitator associated with the Anti-foam System to the Monitoring and Control System (MCS). The 242-A Evaporator is a waste treatment facility designed to reduce liquid waste volumes currently stored in the Hanford Area double shell Waste Storage Tanks. The evaporator uses evaporative concentration to achieve this volume reduction, returning the concentrated slurry to the double-shell tanks for storage and, at the same time, releasing the process effluent to a retention facilities for eventual treatment and release to the environment

  8. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  9. Semiclassical description of hot nuclear systems

    International Nuclear Information System (INIS)

    Brack, M.

    1984-01-01

    We present semiclassical density variational calculations for highly excited nuclear systems. We employ the newly derived functionals tau[rho] and sigma[rho] of the extended Thomas-Fermi (ETF) model, generalized to finite temperatures. Excellent agreement is reached with Hartree-Fock (HF) results. We also calculated the fission barrier of 240 Pu as a function of the nuclear temperature

  10. The Transportable Measurements Facility (TMF) System Description.

    Science.gov (United States)

    1980-05-23

    sites using different antennas, antenna/site characterization, ATCRBS-mode and DABS-mode processor evaluation, and DABS-based ATC and ATARS system...conditions met. 34 TABLE 3 TMF DATA RECORDED DURING EXPERIMENTS By Word Type 1) By Parameter Word No. of Bits Experiment Number 8 Physical Location Number of

  11. Tritium-containment systems: a tradeoff study

    International Nuclear Information System (INIS)

    Folkers, C.L.; Cena, R.J.

    1978-01-01

    Various design parameters are evaluated that affect the performance of tritium-containment systems for fusion reactors. Our study included a review of such parameters as tritium forms, impurities, catalysts, adsorbents, getters, and as low as reasonably achievable principles. We organized these schemes, which can be considered for treating either air or inert atmospheres, so one could easily make orderly choices and tradeoffs for optimum performance. The relationships examined involved purification-system decontamination factors, flow rates, recycling and leakage, and environmental losses

  12. MITLL 2015 Language Recognition Evaluation System Description

    Science.gov (United States)

    2016-01-27

    912 8.18 qsl-rus Russian 2021 37.80 ara-ary Maghrebi 919 46.91 spa-car Carib. Spa. 194 30.59 ara-arz Egyptian 440 97.27 spa-eur Eur. Spa. 366 8.55...qsl-pol Polish 695 32.14 ara-arb MSA 912 8.18 qsl-rus Russian 2021 37.80 ara-ary Maghrebi 919 46.91 spa-car Carib. Spa. 194 30.59 ara-arz Egyptian ...BOTTLENECK I-VECTOR SYSTEM (BNF1) The Deep Neural Network architecture that we used for this system was composed of seven hidden layers. The sixth

  13. Gravity Probe B data system description

    International Nuclear Information System (INIS)

    Bennett, Norman R

    2015-01-01

    The Gravity Probe B data system, developed, integrated, and tested by Lockheed Missiles and Space Company, and later Lockheed Martin Corporation, included flight and ground command, control, and communications software. The development was greatly facilitated, conceptually and by the transfer of key personnel, through Lockheed’s earlier flight and ground test software development for the Hubble Space Telescope (HST). Key design challenges included the tight mission timeline (17 months, 9 days of on-orbit operation), the need to tune the system once on-orbit, and limited 2 Kbps real-time data rates and ground asset availability. The result was a completely integrated space vehicle and Stanford mission operations center, which successfully collected and archived 97% of the ‘guide star valid’ data to support the science analysis. Lessons learned and incorporated from the HST flight software development and on-orbit support experience, and Lockheed’s independent research and development effort, will be discussed. (paper)

  14. Hamiltonian description and quantization of dissipative systems

    Science.gov (United States)

    Enz, Charles P.

    1994-09-01

    Dissipative systems are described by a Hamiltonian, combined with a “dynamical matrix” which generalizes the simplectic form of the equations of motion. Criteria for dissipation are given and the examples of a particle with friction and of the Lotka-Volterra model are presented. Quantization is first introduced by translating generalized Poisson brackets into commutators and anticommutators. Then a generalized Schrödinger equation expressed by a dynamical matrix is constructed and discussed.

  15. Thermionic system evaluated test (TSET) facility description

    Science.gov (United States)

    Fairchild, Jerry F.; Koonmen, James P.; Thome, Frank V.

    1992-01-01

    A consortium of US agencies are involved in the Thermionic System Evaluation Test (TSET) which is being supported by the Strategic Defense Initiative Organization (SDIO). The project is a ground test of an unfueled Soviet TOPAZ-II in-core thermionic space reactor powered by electrical heat. It is part of the United States' national thermionic space nuclear power program. It will be tested in Albuquerque, New Mexico at the New Mexico Engineering Research Institute complex by the Phillips Laboratoty, Sandia National Laboratories, Los Alamos National Laboratory, and the University of New Mexico. One of TSET's many objectives is to demonstrate that the US can operate and test a complete space nuclear power system, in the electrical heater configuration, at a low cost. Great efforts have been made to help reduce facility costs during the first phase of this project. These costs include structural, mechanical, and electrical modifications to the existing facility as well as the installation of additional emergency systems to mitigate the effects of utility power losses and alkali metal fires.

  16. Physical description of nuclear materials identification system (NMIS) signatures

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Mullens, J.A.; Mattingly, J.K.; Valentine, T.E.

    2000-01-01

    This paper describes all time and frequency analysis parameters measured with a new correlation processor (capability up to 1 GHz sampling rates and up to five input data channels) for three input channels: (1) the 252 Cf source ionization chamber; (2) a detection channel; and (3) a second detection channel. An intuitive and physical description of the various measured quantities is given as well as a brief mathematical description and a brief description of how the data are acquired. If the full five-channel capability is used, the number of measured quantities increases in number but not in type. The parameters provided by this new processor can be divided into two general classes: time analysis signatures and their related frequency analysis signatures. The time analysis signatures include the number of time m pulses occurs in a time interval, that is triggered randomly, upon a detection event, or upon a source fission event triggered. From the number of pulses in a time interval, the moments, factorial moments, and Feynmann variance can be obtained. Recent implementations of third- and fourth-order time and frequency analysis signatures in this processor are also briefly described. Thus, this processor used with a timed source of input neutrons contains all of the information from a pulsed neutron measurement, one and two detector Rossi-α measurements, multiplicity measurements, and third- and fourth-order correlation functions. This processor, although originally designed for active measurements with a 252 Cf interrogating source, has been successfully used passively (without 252 Cf source) for systems with inherent neutron sources such as fissile systems of plutonium. Data from active measurements with an 18.75 kg highly enriched uranium (93.2 wt%, 235 U) metal casting for storage are presented to illustrate some of the various time and frequency analysis parameters. This processor, which is a five-channel time correlation analyzer with time channel widths

  17. Physics detector simulation facility system software description

    International Nuclear Information System (INIS)

    Allen, J.; Chang, C.; Estep, P.; Huang, J.; Liu, J.; Marquez, M.; Mestad, S.; Pan, J.; Traversat, B.

    1991-12-01

    Large and costly detectors will be constructed during the next few years to study the interactions produced by the SSC. Efficient, cost-effective designs for these detectors will require careful thought and planning. Because it is not possible to test fully a proposed design in a scaled-down version, the adequacy of a proposed design will be determined by a detailed computer model of the detectors. Physics and detector simulations will be performed on the computer model using high-powered computing system at the Physics Detector Simulation Facility (PDSF). The SSCL has particular computing requirements for high-energy physics (HEP) Monte Carlo calculations for the simulation of SSCL physics and detectors. The numerical calculations to be performed in each simulation are lengthy and detailed; they could require many more months per run on a VAX 11/780 computer and may produce several gigabytes of data per run. Consequently, a distributed computing environment of several networked high-speed computing engines is envisioned to meet these needs. These networked computers will form the basis of a centralized facility for SSCL physics and detector simulation work. Our computer planning groups have determined that the most efficient, cost-effective way to provide these high-performance computing resources at this time is with RISC-based UNIX workstations. The modeling and simulation application software that will run on the computing system is usually written by physicists in FORTRAN language and may need thousands of hours of supercomputing time. The system software is the ''glue'' which integrates the distributed workstations and allows them to be managed as a single entity. This report will address the computing strategy for the SSC

  18. The Norwegian Electric Power System - System Description and Future Developments; Norsk kraftforsyning - dagens system og fremtidig utvikling

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, Janne Merete; Nystuen, Kjell Olav; Fridheim, Haavard; Rutledal, Frode

    2000-09-01

    This report presents a description of the present Norwegian electric power system, as well as a discussion of emerging trends and future developments in this system. The report provides the basis for FFI's current vulnerability analysis of the electric power system. Norway's electric power system is getting increasingly complex, due to a large-scale implementation of electronic components and information systems. Workforce reductions and efficiency improvements dominate the development of the electric power sector. Norway is also becoming increasingly dependent on foreign power sources. These trends provide for an entirely different electric power system than just a few years ago. Also, these trends make it virtually impossible to present a ''static'' description of the system. Thus, the report also contains a scenario, describing possible future developments of the system until 2010. (author)

  19. Relativistic Descriptions of Few-Body Systems

    International Nuclear Information System (INIS)

    Karmanov, V. A.

    2011-01-01

    A brief review of relativistic effects in few-body systems, of theoretical approaches, recent developments and applications is given. Manifestations of relativistic effects in the binding energies, in the electromagnetic form factors and in three-body observables are demonstrated. The three-body forces of relativistic origin are also discussed. We conclude that relativistic effects in nuclei can be important in spite of small binding energy. At high momenta they clearly manifest themselves and are necessary to describe the deuteron e.m. form factors. At the same time, there is still a discrepancy in three-body observables which might be a result of less clarity in understanding the corresponding relativistic effects, the relativistic NN kernel and the three-body forces. Relativistic few-body physics remains to be a field of very intensive and fruitful researches. (author)

  20. On the Lagrangian description of dissipative systems

    Science.gov (United States)

    Martínez-Pérez, N. E.; Ramírez, C.

    2018-03-01

    We consider the Lagrangian formulation with duplicated variables of dissipative mechanical systems. The application of Noether theorem leads to physical observable quantities which are not conserved, like energy and angular momentum, and conserved quantities, like the Hamiltonian, that generate symmetry transformations and do not correspond to observables. We show that there are simple relations among the equations satisfied by these two types of quantities. In the case of the damped harmonic oscillator, from the quantities obtained by the Noether theorem follows the algebra of Feshbach and Tikochinsky. Furthermore, if we consider the whole dynamics, the degrees of freedom separate into a physical and an unphysical sector. We analyze several cases, with linear and nonlinear dissipative forces; the physical consistency of the solutions is ensured, observing that the unphysical sector has always the trivial solution.

  1. Demonstration of an Emergency Containment System

    International Nuclear Information System (INIS)

    Flanagan, T.M.; Rogers, M.L.; Wilkes, W.R.

    1978-01-01

    A system called an Emergency Containment System (ECS) to be used for tertiary containment of tritium was reported at the 13th Air Cleaning Conference. This system was part of the Tritium Effluent Control Laboratory then under construction at Mound Facility. A series of experiments has recently been conducted to evaluate the performance of an ECS in capturing tritium accidentally released into an operating laboratory. The ECS is an automatically actuated laboratory air detritiation system utilizing a catalytic oxidation reactor and presaturated oxide adsorption/exchange columns. In the event of an accidental release of tritium into the laboratory, the ECS is automatically activated, and quick-acting pneumatic dampers divert the laboratory air supply and exhaust through the ECS until room concentrations are returned to safe operating levels. The results of the experiments have shown that a tertiary containment of tritium is feasible. In the event of a catastrophic accident, the ECS is capable of preventing the release of a large quantity of tritium to the environment

  2. Detector correction in large container inspection systems

    CERN Document Server

    Kang Ke Jun; Chen Zhi Qiang

    2002-01-01

    In large container inspection systems, the image is constructed by parallel scanning with a one-dimensional detector array with a linac used as the X-ray source. The linear nonuniformity and nonlinearity of multiple detectors and the nonuniform intensity distribution of the X-ray sector beam result in horizontal striations in the scan image. This greatly impairs the image quality, so the image needs to be corrected. The correction parameters are determined experimentally by scaling the detector responses at multiple points with logarithm interpolation of the results. The horizontal striations are eliminated by modifying the original image data with the correction parameters. This method has proven to be effective and applicable in large container inspection systems

  3. Halon 1301 protection system for nuclear containments

    International Nuclear Information System (INIS)

    McHale, E.T.

    1981-01-01

    Halon 1301 can provide protection against any combustion hazard that hydrogen gas might present in an LWR containment following a loss-of-coolant accident. A development program was conducted, comprising analytical study, laboratory experiments and large-scale testing, to define the requirements for a Halon 1301 system and to examine certain operational problems that were hypothesized. Some results of the study are presented in this paper

  4. Standard-D hydrogen monitoring system, system design description

    International Nuclear Information System (INIS)

    Schneider, T.C.

    1996-01-01

    During most of the year, it is assumed that the vapor space in the 177 radioactive waste tanks on the Hanford Project site contain a uniform mixture of gases. Several of these waste tanks (currently twenty-five, 6 Double Shell Tanks and 19 Single Shell Tanks) were identified as having the potential for the buildup of gasses to a flammable level. An active ventilation system in the Double Shell Tanks and a passive ventilation system in the Single Shell Tanks provides a method of expelling gasses from the tanks. A gas release from a tank causes a temporary rise in the tank pressure, and a potential for increased concentration of hydrogen gas in the vapor space. The gas is released via the ventilation systems until a uniform gas mixture in the vapor space is once again achieved. The Standard Hydrogen Monitoring System (SHMS) is designed to monitor and quantify the percent hydrogen concentration during these potential gas releases. This document describes the design of the Standard-D Hydrogen Monitoring System, (SHMS-D) and its components as it differs from the original SHMS

  5. Standard-B Hydrogen Monitoring System, system design description

    International Nuclear Information System (INIS)

    Schneider, T.C.

    1995-01-01

    During most of the year, it is assumed that the vapor in the 177 radioactive waste tanks on the Hanford Project site contain a uniform mixture of gases. Several of these waste tanks (currently twenty five, 6 Double Shell Tanks and 19 Single Shell Tanks) were identified as having the potential for the buildup of gases to a flammable level. An active ventilation system in the Double Shell Tanks and a passive ventilation system in the Single Shell Tanks provides a method of expelling gases from the tanks. A gas release from a tank causes a temporary rise in the tank pressure, and a potential for increased concentration of hydrogen gas in the vapor space. The gas is released via the ventilation systems until a uniform gas mixture in the vapor space is once again achieved. This document describes the design of the Standard-B Hydrogen Monitoring System, (SHMS) and its components as it differs from the original SHMS. The differences are derived from changes made to improve the system performance but not implemented in all the installed enclosures

  6. Explosion testing for the container venting system

    International Nuclear Information System (INIS)

    Cashdollar, K.L.; Green, G.M.; Thomas, R.A.; Demiter, J.A.

    1993-01-01

    As part of the study of the hazards of inspecting nuclear waste stored at the Hanford Site, the US Department of Energy and Westinghouse Hanford Company have developed a container venting system to sample the gases that may be present in various metal drums and other containers. In support of this work, the US Bureau of Mines has studied the probability of ignition while drilling into drums and other containers that may contain flammable gas mixtures. The Westinghouse Hanford Company drilling procedure was simulated by tests conducted in the Bureau's 8-liter chamber, using the same type of pneumatic drill that will be used at the Hanford Site. There were no ignitions of near-stoichiometric hydrogen-air or methane-air mixtures during the drilling tests. The temperatures of the drill bits and lids were measured by an infrared video camera during the drilling tests. These measured temperatures are significantly lower than the ∼500 degree C autoignition temperature of uniformly heated hydrogen-air or the ∼600 degree C autoignition temperature of uniformly heated methane-air. The temperatures are substantially lower than the 750 degree C ignition temperature of hydrogen-air and 1,220 degree C temperature of methane-air when heated by a 1-m-diameter wire

  7. The AutoBayes Program Synthesis System: System Description

    Science.gov (United States)

    Fischer, Bernd; Pressburger, Thomas; Rosu, Grigore; Schumann, Johann; Norvog, Peter (Technical Monitor)

    2001-01-01

    AUTOBAYES is a fully automatic program synthesis system for the statistical data analysis domain. Its input is a concise description of a data analysis problem in the form of a statistical model; its output is optimized and fully documented C/C++ code which can be linked dynamically into the Matlab and Octave environments. AUTOBAYES synthesizes code by a schema-guided deductive process. Schemas (i.e., code templates with associated semantic constraints) are applied to the original problem and recursively to emerging subproblems. AUTOBAYES complements this approach by symbolic computation to derive closed-form solutions whenever possible. In this paper, we concentrate on the interaction between the symbolic computations and the deductive synthesis process. A statistical model specifies for each problem variable (i.e., data or parameter) its properties and dependencies in the form of a probability distribution, A typical data analysis task is to estimate the best possible parameter values from the given observations or measurements. The following example models normal-distributed data but takes prior information (e.g., from previous experiments) on the data's mean value and variance into account.

  8. 100-N technical manual. Volume 2A: Systems descriptions

    Energy Technology Data Exchange (ETDEWEB)

    1963-12-31

    This report contains engineering drawings for the control room, reactor monitoring systems, and reactor control systems for the N reactor. Each console in the control room is detailed. Other systems discussed are: stack air monitoring system, charging machine control systems, and heating and ventilation control systems. A N reactor plant glossary is included.

  9. Irreducible descriptive sets of attributes for information systems

    KAUST Repository

    Moshkov, Mikhail

    2010-01-01

    The maximal consistent extension Ext(S) of a given information system S consists of all objects corresponding to attribute values from S which are consistent with all true and realizable rules extracted from the original information system S. An irreducible descriptive set for the considered information system S is a minimal (relative to the inclusion) set B of attributes which defines exactly the set Ext(S) by means of true and realizable rules constructed over attributes from the considered set B. We show that there exists only one irreducible descriptive set of attributes. We present a polynomial algorithm for this set construction. We also study relationships between the cardinality of irreducible descriptive set of attributes and the number of attributes in S. The obtained results will be useful for the design of concurrent data models from experimental data. © 2010 Springer-Verlag.

  10. Cold Vacuum Drying facility civil - structural system design description (SYS 06)

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying (CVD) Facility civil - structural system. This system consists of the facility structure, including the administrative and process areas. The system's primary purpose is to provide for a facility to house the CVD process and personnel and to provide a tertiary level of containment. The document provides a description of the facility and demonstrates how the design meets the various requirements imposed by the safety analysis report and the design requirements document

  11. GOTHIC Simulation of Passive Containment Cooling System

    International Nuclear Information System (INIS)

    Ha, Huiun; Kim, Hangon

    2013-01-01

    The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR

  12. GOTHIC Simulation of Passive Containment Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Kim, Hangon [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR.

  13. On the description of fault-tolerant systems

    International Nuclear Information System (INIS)

    Syrbe, M.

    1980-01-01

    Various demands by increasing complexity and the disposability of new technologies, like the One-chip-microcomputer and fiber optics, lead to control systems, which are built as decentralized distributed multi-microcomputersystems. They realize not only new control functions but they also open possibilities to increase availability by fault-tolerance. The design or the selection and lay-out of such systems require a quantitative description of these systems. This is possible on the bases of the set of hardware and software moduls of the system by the use of queuing models, reliability nets and diagnostic graphs. This is shown by an example of a practically applied Really Distributed Computer Control System (RDC-System). Computer aided methods for these system descriptions are emphasized. (orig.) [de

  14. SWR 1000 related containment cooling system tests in PANDA

    International Nuclear Information System (INIS)

    Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

    2000-01-01

    Since 1991 the Paul Scherrer Institute has participated in the investigations of several of the new passive Advanced Light Water Reactor designs proposed world-wide. The current phase of the project, ALPHA-II, is focused on both the boiling water and the pressurized water reactor passive designs and consists of three projects under the sponsorship of the European Commission. The paper describes the performed PANDA transient system tests related to one of these projects, called 'BWR R and D Cluster for Innovative Passive Safety Systems (IPSS)', and details the PSI contribution to the experimental investigation of passive containment cooling by a Building Condenser system which is part of the advanced Boiling Water Reactor SWR 1000 designed by Siemens. First, a short description of the relevant systems of the SWR 1000 design and its simulation in the PANDA facility are presented. After the description of the experimental programme for the large-scale integral system test investigations in the PANDA facility, the main results of the performed tests are also given. Finally, the main conclusions, based on the to date available experimental results and their analysis, are summarised. (author)

  15. Database management system for large container inspection system

    International Nuclear Information System (INIS)

    Gao Wenhuan; Li Zheng; Kang Kejun; Song Binshan; Liu Fang

    1998-01-01

    Large Container Inspection System (LCIS) based on radiation imaging technology is a powerful tool for the Customs to check the contents inside a large container without opening it. The author has discussed a database application system, as a part of Signal and Image System (SIS), for the LCIS. The basic requirements analysis was done first. Then the selections of computer hardware, operating system, and database management system were made according to the technology and market products circumstance. Based on the above considerations, a database application system with central management and distributed operation features has been implemented

  16. Conceptual modular description of the high-level waste management system for system studies model development

    International Nuclear Information System (INIS)

    McKee, R.W.; Young, J.R.; Konzek, G.J.

    1992-08-01

    This document presents modular descriptions of possible alternative components of the federal high-level radioactive waste management system and the procedures for combining these modules to obtain descriptions for alternative configurations of that system. The 20 separate system component modules presented here can be combined to obtain a description of any of the 17 alternative system configurations (i.e., scenarios) that were evaluated in the MRS Systems Studies program (DOE 1989a). First-approximation descriptions of other yet-undefined system configurations could also be developed for system study purposes from this database. The descriptions include, in a modular format, both functional descriptions of the processes in the waste management system, plus physical descriptions of the equipment and facilities necessary for performance of those functions

  17. Excited state characterization of carbonyl containing carotenoids: a comparison between single and multireference descriptions

    Science.gov (United States)

    Spezia, Riccardo; Knecht, Stefan; Mennucci, Benedetta

    Carotenoids can play multiple roles in biological photoreceptors thanks to their rich photophysics. In the present work, we have investigated six of the most common carbonyl containing carotenoids: Echinenone, Canthaxanthin, Astaxanthin, Fucoxanthin, Capsanthin and Capsorubin. Their excitation properties are investigated by means of a hybrid density functional theory (DFT) and multireference configuration interaction (MRCI) approach to elucidate the role of the carbonyl group: the bright transition is of {\\pi}{\\pi}* character, as expected, but the presence of a C=O moiety reduces the energy of n{\\pi}* transitions which may become closer to the {\\pi}{\\pi}* transition, in particular as the conjugation chain decreases. This can be related to the presence of a low-lying charge transfer state typical of short carbonyl- containing carotenoids. The DFT/MRCI results are finally used to benchmark single- reference time-dependent DFT-based methods: among the investigated functionals, the meta- GGA (and in particular M11L and MN12L) functionals show to perform the best for all six investigated systems.

  18. Cold Vacuum Drying (CVD) Electrical System Design Description

    International Nuclear Information System (INIS)

    SINGH, G.

    2000-01-01

    This system design description (SDD) provides a technical explanation of the design and operation of the electrical system for the Cold Vacuum Drying Facility (CVDF). This SDD also identifies the requirements, and the basis for the requirements and details on how the requirements have been implemented in the design and construction of the facility. This SDD also provides general guidance for the surveillance, testing, and maintenance of this system

  19. Surface system Forsmark. Site descriptive modelling SDM-Site Forsmark

    International Nuclear Information System (INIS)

    Lindborg, Tobias

    2008-12-01

    % of some of the delineated sub-catchments. No major water courses flow through the central part of the site investigation area. Many brooks in the area have been deepened for considerable distances for drainage purposes. The horizontal hydraulic conductivity and specific yield of the till, the values of which are based on measurements, are typical or slightly higher than in the surrounding region. Groundwater levels in Quaternary deposits are very shallow, on average less than 0.7 m below ground during 50% of the time. Post-glacial land uplift, in combination with the flat topography, implies fast shoreline displacement. This has resulted in a young terrestrial system that contains a number of newborn lakes and wetlands. The recently isolated and shallow oligotrophic hardwater lakes that are typical for the area are unique in Sweden.The marine ecosystem at Forsmark is situated in a relatively productive coastal area in a region of otherwise fairly low primary production. The seabed is dominated by erosion and transport bottoms with heterogeneous and mobile sediments, consisting mainly of sand and gravel with varying fractions of glacial clay. Based on an overall conceptual model, it was possible to identify pools and fluxes of elements in the landscape that are of potential relevance for a safety assessment. The quantification of these elements, using both field- and model-based estimates, makes it possible to determine the relative importance of the different ecosystems with regard to elemental transport and accumulation. A special emphasis has been put on the description of transport and accumulation of organic matter, since detailed knowledge on the carbon dynamics provides a way of analysing how different ecosystem components are linked to each other through fluxes of energy, i.e. carbon. This provides a baseline for making predictions of dispersal and accumulation of matter, including radionuclides, within and between ecosystems. By this approach, the safety

  20. Green-function description of dense polymeric systems

    NARCIS (Netherlands)

    Schoot, van der P.P.A.M.

    2000-01-01

    A self-consistent Green-function description of concentrated polymer solutions and dense polymeric melts is presented. The method, which applies to both uniform and nonuniform systems, is used in this work to calculate the static structure factor of a homogeneous fluid of Gaussian model chains.

  1. Description of disintegration in a three-body system

    CERN Document Server

    Takibaev, N Z

    2000-01-01

    In the frame of approach based on the effective potential of interaction between constituents, description of inelastic transition, in particularly, the processes of system disintegration. Relationship is shown between the approach results and those of the theory of final state interaction where coefficients of reaction gaining factor are determined. (author)

  2. 44 CFR 334.5 - GMR system description.

    Science.gov (United States)

    2010-10-01

    ... problems. (b) Stage 2, Crisis Management. During the crisis management stage, GMR plans are reviewed and... developed). For example, a Federal department or agency might divide “Crisis Management” into two, three, or... 44 Emergency Management and Assistance 1 2010-10-01 2010-10-01 false GMR system description. 334.5...

  3. System description: Isabelle/jEdit in 2014

    Directory of Open Access Journals (Sweden)

    Makarius Wenzel

    2014-10-01

    Full Text Available This is an updated system description for Isabelle/jEdit, according to the official release Isabelle2014 (August 2014. The following new PIDE concepts are explained: asynchronous print functions and document overlays, syntactic and semantic completion, editor navigation, management of auxiliary files within the document-model.

  4. A Petri Net Definition of a System Description Language

    DEFF Research Database (Denmark)

    Jensen, Kurt; Kyng, Morten; Madsen, Ole Lehrmann

    1979-01-01

    This paper introduces a language for the description of systems with concurrency, and presents a formal definition of its semantics. The language is based on Delta and the semantic model is an extension of Petri nets with a data part and with expressions attached to transitions and to places....

  5. System design description for master equipment list, phase I

    International Nuclear Information System (INIS)

    Sandoval, J.D.

    1997-01-01

    This System Design Description (SDD) is for the Master Equipment List Phase I (MEL). It has been prepared following the WI-IC-CM-3-10, ''Software Practices,'' (Ref. 6). This SDD describes the internal design for implementation of the MEL Phase I

  6. Irreducible descriptive sets of attributes for information systems

    KAUST Repository

    Moshkov, Mikhail; Skowron, Andrzej; Suraj, Zbigniew

    2010-01-01

    . An irreducible descriptive set for the considered information system S is a minimal (relative to the inclusion) set B of attributes which defines exactly the set Ext(S) by means of true and realizable rules constructed over attributes from the considered set B

  7. System design description for sampling fuel in K basins

    International Nuclear Information System (INIS)

    Baker, R.B.

    1996-01-01

    This System Design Description provides: (1) statements of the Spent Nuclear Fuel Projects (SNFP) needs requiring sampling of fuel in the K East and K West Basins, (2) the sampling equipment functions and requirements, (3) a general work plan and the design logic being followed to develop the equipment, and (4) a summary description of the design for the sampling equipment. The report summarizes the integrated application of both the subject equipment and the canister sludge sampler in near-term characterization campaigns at K Basins

  8. Fire protection countermeasures for containment ventilation systems

    International Nuclear Information System (INIS)

    Alvares, N.; Beason, D.; Bergman, V.; Creighton, J.; Ford, H.; Lipska, A.

    1980-01-01

    The goal of this project is to find countermeasures to protect High Efficiency Particulate Air (HEPA) filters, in exit ventilation ducts, from the heat and smoke generated by fire. Initially, methods were developed to cool fire-heated air by fine water spray upstream of the filters. It was recognized that smoke aerosol exposure to HEPA filters could also cause disruption of the containment system. Through testing and analysis, several methods to partially mitigate the smoke exposure to the HEPA filters were identified. A continuous, movable, high-efficiency prefilter using modified commercial equipment was designed. The technique is capable of protecting HEPA filters over the total time duration of the test fires. The reason for success involved the modification of the prefiltration media. Commercially available filter media has particle sorption efficiency that is inversely proportional to media strength. To achieve properties of both efficiency and strength, rolling filter media were laminated with the desired properties. The approach was Edisonian, but truncation in short order to a combination of prefilters was effective. The application of this technique was qualified, since it is of use only to protect HEPA filters from fire-generated smoke aerosols. It is not believed that this technique is cost effective in the total spectrum of containment systems, especially if standard fire protection systems are available in the space. But in areas of high-fire risk, where the potential fuel load is large and ignition sources are plentiful, the complication of a rolling prefilter in exit ventilation ducts to protect HEPA filters from smoke aerosols is definitely justified

  9. 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...

  10. Preparation of plant and system design description documents

    International Nuclear Information System (INIS)

    1989-01-01

    This standard prescribes the purpose, scope, organization, and content of plant design requirements (PDR) documents and system design descriptions (SDDs), to provide a unified approach to their preparation and use by a project as the principal means to establish the plant design requirements and to establish, describe, and control the individual system designs from conception and throughout the lifetime of the plant. The Electric Power Research Institute's Advanced Light Water Reactor (LWR) Requirements Document should be considered for LWR plants

  11. Buried waste containment system materials. Final Report

    International Nuclear Information System (INIS)

    Weidner, J.R.; Shaw, P.G.

    1997-10-01

    This report describes the results of a test program to validate the application of a latex-modified cement formulation for use with the Buried Waste Containment System (BWCS) process during a proof of principle (POP) demonstration. The test program included three objectives. One objective was to validate the barrier material mix formulation to be used with the BWCS equipment. A basic mix formula for initial trials was supplied by the cement and latex vendors. The suitability of the material for BWCS application was verified by laboratory testing at the Idaho National Engineering and Environmental Laboratory (INEEL). A second objective was to determine if the POP BWCS material emplacement process adversely affected the barrier material properties. This objective was met by measuring and comparing properties of material prepared in the INEEL Materials Testing Laboratory (MTL) with identical properties of material produced by the BWCS field tests. These measurements included hydraulic conductivity to determine if the material met the US Environmental Protection Agency (EPA) requirements for barriers used for hazardous waste sites, petrographic analysis to allow an assessment of barrier material separation and segregation during emplacement, and a set of mechanical property tests typical of concrete characterization. The third objective was to measure the hydraulic properties of barrier material containing a stop-start joint to determine if such a feature would meet the EPA requirements for hazardous waste site barriers

  12. Licensee Event Report system: description of system and guidelines for reporting

    International Nuclear Information System (INIS)

    Hebdon, F.J.

    1983-09-01

    On July 26, 1983, the Commission published in the Federal Register a final rule (10 CFR 50.73) that modifies and codifies the Licensee Event Report (LER) system. The rule becomes effective on January 1, 1984. This NUREG provides supporting information and guidance that will be of interest to persons responsible for the preparation and review of LERs. The information contained in this NUREG includes: (1) a brief description of how LERs are analyzed by the NRC; (2) a restatement of the guidance contained in the Statement of Consideration that accompanied the publication of the LER rule; (3) a set of examples of potentially reportable events with staff comments on the actual reportability of each event; (4) guidance on how to prepare an LER, including the LER forms; and (5) guidance on submittal of LERs

  13. Positron-containing systems and positron diagnostics

    International Nuclear Information System (INIS)

    1978-01-01

    The results of the experimental and theoretical investigations are presented. Considered are quantum-mechanical calculations of wave functions describing the states of positron-containing atomic systems and of cross-sections of the processes characterizing different interactions, and also the calculations of the behaviour of positrons in gases in the presence of an electric field. The results of experimental tests are presented by the data describing the behaviour of positrons and positronium in liquids, polymers and elastomers, complex oxides and in different solids. New equipment and systems developed on the basis of current studies are described. Examined is a possibility of applying the methods of model and effective potentials for studying the bound states of positron systems and for calculating cross-sections of elementary processes of elastic and inelastic collisions with a positron involved. The experimental works described indicate new possibilities of the positron diagnosis method: investigation of thin layers and films of semiconductor materials, defining the nature of chemical bonds in semiconductors, determination of the dislocation density in deformed semiconductors, derivation of important quantitative information of the energy states of radiation defects in them

  14. WRAP Module 1 data management system (DMS) software design description (SDD)

    International Nuclear Information System (INIS)

    Weidert, J.R.

    1996-01-01

    Revision 2 of the Waste Receiving and Processing (WRAP) Module 1 Data Management System (DMS) Preliminary Software Design Description (PSDD) provides a high-level design description of the system. The WRAP 1 DMS is required to collect, store, and report data related to certification, tracking, packaging, repackaging, processing, and shipment of waste processed or stored at the WRAP 1 facility. The WRAP 1 DMS SDD is used as the primary medium for communication software design information. This release provides design descriptions for the following process modules produced under Phase 1 of the development effort: Receiving Drum or Box Containers Process Routing and Picklists; Waste Inventory by Location and/or Container Relationships; LLW Process Glovebox Facility Radiologic Material Inventory Check (partial); Shipping (partial production); Drum or Box NDE Operations; and Drum or Box NDA Operations Data Review (partial production). In addition, design descriptions are included for the following process modules scheduled for development under Phases 2 and 3: Activity Comment; LLW RWM Glovebox Sample Management; TRU Process Glovebox; TRU RWM Glovebox; and TRUPACT Processing. Detailed design descriptions for Reports and Facility Metrics have also been provided for in Revision 2 of this document

  15. Rock Visualization System. Technical description (RVS v.3.5)

    Energy Technology Data Exchange (ETDEWEB)

    Curtis, P.; Elfstroem, M.; Markstroem, I. [FB Engineering, Goeteborg (Sweden)

    2004-03-01

    The Rock Visualization System (RVS) has been developed by SKB for use in visualizing geological and engineering data in 3D. The purpose of this report is to provide a technical description of RVS aimed at potential program users and interested parties as well as fulfilling the function of a more general RVS reference that can be cited when writing other technical reports. It is a description of RVS version 3.5. Updated versions of this report or addenda will be made available following further development of RVS and the release of subsequent versions of the program. The report covers the following main items: Technical description of the program with illustrations and examples; Limitations of the program and of functionality. For most RVS functions step-by-step tutorials are available describing how a particular function can be used to carryout a specific task. A complete set of updated tutorials is issued with each new version release of the RVS program. However, the tutorials do not cover all the possible uses of all the individual functions but rather give an overall view of their functionality. A detailed description of every RVS function and how it can be used is included in the RVS online Help system.

  16. Rock Visualization System. Technical description (RVS v.3.5)

    International Nuclear Information System (INIS)

    Curtis, P.; Elfstroem, M.; Markstroem, I.

    2004-03-01

    The Rock Visualization System (RVS) has been developed by SKB for use in visualizing geological and engineering data in 3D. The purpose of this report is to provide a technical description of RVS aimed at potential program users and interested parties as well as fulfilling the function of a more general RVS reference that can be cited when writing other technical reports. It is a description of RVS version 3.5. Updated versions of this report or addenda will be made available following further development of RVS and the release of subsequent versions of the program. The report covers the following main items: Technical description of the program with illustrations and examples; Limitations of the program and of functionality. For most RVS functions step-by-step tutorials are available describing how a particular function can be used to carryout a specific task. A complete set of updated tutorials is issued with each new version release of the RVS program. However, the tutorials do not cover all the possible uses of all the individual functions but rather give an overall view of their functionality. A detailed description of every RVS function and how it can be used is included in the RVS online Help system

  17. Tracer verification and monitoring of containment systems

    International Nuclear Information System (INIS)

    Lowry, W.; Dunn, S.D.; Williams, C.

    1996-01-01

    In-situ barrier emplacement techniques and materials for the containment of high-risk contaminants in soils are currently being developed by the Department of Energy (DOE). Because of their relatively high cost, the barriers are intended to be used in cases where the risk is too great to remove the contaminants, the contaminants are too difficult to remove with current technologies, or the potential for movement of the contaminants to the water table is so high that immediate action needs to be taken to reduce health risks. Consequently, barriers are primarily intended for use in high-risk sites where few viable alternatives exist to stop the movement of contaminants in the near term. Assessing the integrity of the barrier once it is emplaced, and during its anticipated life, is a very difficult but necessary requirement. Existing surface-based and borehole geophysical techniques do not provide the degree of resolution required to assure the formation of an integral in-situ barrier. Science and Engineering Associates, Inc., (SEA) and Sandia National Laboratories (SNL) are developing a quantitative subsurface barrier assessment system using gaseous tracers. Called SEAtrace trademark, this system integrates an autonomous, multipoint soil vapor sampling and analysis system with a global optimization modeling methodology to pinpoint leak sources and sizes in real time. SEAtrace trademark is applicable to impermeable barrier emplacements above the water table, providing a conservative assessment of barrier integrity after emplacement, as well as a long term integrity monitoring function. The SEAtrace trademark system is being developed under funding by the DOE-EM Subsurface Contaminant Focus Area

  18. WRAP Module 1 data management system (DMS) software design description (SDD)

    Energy Technology Data Exchange (ETDEWEB)

    Talmage, P.A.

    1995-03-17

    The Waste Receiving and Processing (WRAP) Module 1 Data Management System (DMS) System Design Description (SDD) describes the logical and physical architecture of the system. The WRAP 1 DMS SDD formally partitions the elements of the system described in the WRAP 1 DMS Software requirements specification into design objects and describes the key properties and relationships among the design objects and interfaces with external systems such as the WRAP Plant Control System (PCS). The WRAP 1 DMS SDD can be thought of as a detailed blueprint for implementation activities. The design descriptions contained within this document will describe, in detail, the software products that will be developed to assist the Project W-026, Waste Receiving and Processing Module 1, in their management functions. The WRAP 1 DMS is required to collect, store, and report data related to certification, tracking, packaging, repackaging, processing, and shipment of waste processed or stored at the WRAP 1 facility.

  19. WRAP Module 1 data management system (DMS) software design description (SDD)

    International Nuclear Information System (INIS)

    Talmage, P.A.

    1995-01-01

    The Waste Receiving and Processing (WRAP) Module 1 Data Management System (DMS) System Design Description (SDD) describes the logical and physical architecture of the system. The WRAP 1 DMS SDD formally partitions the elements of the system described in the WRAP 1 DMS Software requirements specification into design objects and describes the key properties and relationships among the design objects and interfaces with external systems such as the WRAP Plant Control System (PCS). The WRAP 1 DMS SDD can be thought of as a detailed blueprint for implementation activities. The design descriptions contained within this document will describe, in detail, the software products that will be developed to assist the Project W-026, Waste Receiving and Processing Module 1, in their management functions. The WRAP 1 DMS is required to collect, store, and report data related to certification, tracking, packaging, repackaging, processing, and shipment of waste processed or stored at the WRAP 1 facility

  20. Description of the control and safety systems of the RA reactor

    International Nuclear Information System (INIS)

    Popovic, B.; Pesic, M.

    1962-01-01

    This report contains detailed description and scheme of the control and safety system of the RA reactor. It consists of interconnected five systems: for automated regulation; compensation rods; safety rods; power density measurement device; period meter; automated D 2 O level meter in the core. Automated regulation system is divided into two parts: basic system for reactor operation regime at power from 10kW - 10 MW and precise regulation system for operation at set-up power level up to 10 kW which is used occasionally

  1. US Department of Energy, Richland Operations Office Integrated Safety Management System Program Description

    International Nuclear Information System (INIS)

    SHOOP, D.S.

    2000-01-01

    The purpose of this Integrated Safety Management System (ISMS) Program Description (PD) is to describe the U.S. Department of Energy (DOE), Richland Operations Office (RL) ISMS as implemented through the RL Integrated Management System (RIMS). This PD does not impose additional requirements but rather provides an overview describing how various parts of the ISMS fit together. Specific requirements for each of the core functions and guiding principles are established in other implementing processes, procedures, and program descriptions that comprise RIMS. RL is organized to conduct work through operating contracts; therefore, it is extremely difficult to provide an adequate ISMS description that only addresses RL functions. Of necessity, this PD contains some information on contractor processes and procedures which then require RL approval or oversight

  2. Performance assessment of containment filtered venting system with Venturi scrubber

    International Nuclear Information System (INIS)

    Adinarayna, K.N.V.; Ali, Seik Mansoor; Balasubramaniyan, V.

    2015-01-01

    Venting through appropriate filtration systems is now being considered as a severe accident management strategy for maintaining the containment integrity and also as a means to reduce the radiological consequences to the public and environment. The option of filtered containment venting appears to have assumed significance in the post- Fukushima accident backdrop. Back-fitting of a suitable Venturi scrubber based CFVS for the Indian BWRs (TAPS- 1 and 2) at Tarapur is now being contemplated. Several key issues need to be carefully addressed for ensuring the desired functional capability of such a system. At the outset, this paper highlights a few thermal hydraulic issues that are of interest from regulatory perspective. This is followed by a detailed description of the mathematical models developed for assessing the depressurization characteristics of CFVS, energy absorption capacity of the Scrubber Tank (ST) water inventory, iodine removal and aerosol retention capability etc. Finally, application of these models to investigate the response of CFVS under twin unit SBO conditions in TAPS-1 and 2 is presented. The studies presented here give insight into the key variables affecting the CFVS performance and would be useful to both the system designer as well as the regulator. (author)

  3. Architecture Descriptions. A Contribution to Modeling of Production System Architecture

    DEFF Research Database (Denmark)

    Jepsen, Allan Dam; Hvam, Lars

    a proper understanding of the architecture phenomenon and the ability to describe it in a manner that allow the architecture to be communicated to and handled by stakeholders throughout the company. Despite the existence of several design philosophies in production system design such as Lean, that focus...... a diverse set of stakeholder domains and tools in the production system life cycle. To support such activities, a contribution is made to the identification and referencing of production system elements within architecture descriptions as part of the reference architecture framework. The contribution...

  4. Design description of the Tangaye Village photovoltaic power system

    Science.gov (United States)

    Martz, J. E.; Ratajczak, A. F.

    1982-01-01

    The engineering design of a stand alone photovoltaic (PV) powered grain mill and water pump for the village of Tangaye, Upper Volta is described. The socioeconomic effects of reducing the time required by women in rural areas for drawing water and grinding grain were studied. The suitability of photovoltaic technology for use in rural areas by people of limited technical training was demonstrated. The PV system consists of a 1.8-kW (peak) solar cell array, 540 ampere hours of battery storage, instrumentation, automatic controls, and a data collection and storage system. The PV system is situated near an improved village well and supplies d.c. power to a grain mill and a water pump. The array is located in a fenced area and the mill, battery, instruments, controls, and data system are in a mill building. A water storage tank is located near the well. The system employs automatic controls which provide battery charge regulation and system over and under voltage protection. This report includes descriptions of the engineering design of the system and of the load that it serves; a discussion of PV array and battery sizing methodology; descriptions of the mechanical and electrical designs including the array, battery, controls, and instrumentation; and a discussion of the safety features. The system became operational on March 1, 1979.

  5. Model and information abstraction for description-driven systems

    International Nuclear Information System (INIS)

    Estrella, F.; McClatchey, R.; Kovacs, Z.; Goff, J.-M.L.

    2001-01-01

    A crucial factor in the creation of adaptable systems dealing with changing requirements is the suitability of the underlying technology in allowing the evolution of the system. A reflective system utilizes an open architecture where implicit system aspects are reified to become explicit first-class (meta-data) objects. These implicit system aspects are often fundamental structures which are inaccessible and immutable, and their reification as meta-data objects can serve as the basis for changes and extensions to the system, making it self-describing. To address the evolvability issue, the author proposes a reflective architecture based on two orthogonal abstractions-model abstraction and information abstraction. In this architecture the modeling abstractions allow for the separation of the description meta-data from the system aspects they represent so that they can be managed and versioned independently, asynchronously and explicitly

  6. Cold Vacuum Drying Instrument Air System Design Description (SYS 12)

    Energy Technology Data Exchange (ETDEWEB)

    SHAPLEY, B.J.; TRAN, Y.S.

    2000-06-05

    This system design description (SDD) addresses the instrument air (IA) system of the spent nuclear fuel (SNF). This IA system provides instrument quality air to the Cold Vacuum Drying (CVD) Facility. The IA system is a general service system that supports the operation of the heating, ventilation, and air conditioning (HVAC) system, the process equipment skids, and process instruments in the CVD Facility. The following discussion is limited to the compressor, dryer, piping, and valving that provide the IA as shown in Drawings H-1-82222, Cold Vacuum Drying Facility Mechanical Utilities Compressed & Instrument Air P&ID, and H-1.82161, Cold Vacuum Drying Facility Process Equipment Skid P&ID MCO/Cusk Interface. Figure 1-1 shows the physical location of the 1A system in the CVD Facility.

  7. The Types of Personal Networks in the Texts Containing Descriptions of Dematerialization of a Subject

    Directory of Open Access Journals (Sweden)

    Ella V. Nesterik

    2015-01-01

    Full Text Available The article examines the types of personal networks found in the descriptions of dematerialization of a subject and reveals the role of the linguistic means expressing the category of personality in the linguistic embodiment of this phenomenon. The research is conducted at the junction of several disciplines, among which text linguistics takes the leading place. The authors come to the conclusion that dematerialization is formed in a literary text by explicit means of expression of personality – predicates of a certain type and pronominal personal network.

  8. Design of the containment spray system

    International Nuclear Information System (INIS)

    1985-12-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The present RFS defines the functional requirements of the containment spray system and proposes certain complementary criteria or methods to be used in its equipment design

  9. Fire protection countermeasures for containment ventilation systems

    International Nuclear Information System (INIS)

    Alvares, N.J.; Beason, D.G.; Bergman, W.; Ford, H.W.; Lipska, A.E.

    1980-01-01

    The goal of this project is to find countermeasures to protect HEPA filters in exit ventilation ducts from the heat and smoke generated by fire. Several methods for partially mitigating the smoke exposure to the HEPA filters were identified through testing and analysis. These independently involve controlling the fuel, controlling the fire, and intercepting the smoke aerosol prior to its sorption on the HEPA filter. Exit duct treatment of aerosols is not unusual in industrial applications and involves the use of scrubbers, prefilters, and inertial impaction, depending on the size, distribution, and concentration of the subject aerosol. However, when these unmodified techniques were applied to smoke aerosols from fires on materials, common to experimental laboratories of LLNL, it was found they offered minimal protection to the HEPA filters. Ultimately, a continuous, movable, high-efficiency prefilter using modified commercial equipment was designed. This technique is capable of protecting HEPA filters over the total duration of the test fires. The reason for success involved the modificaton of the prefiltration media. Commercially available filter media has a particle sorption efficiency that is inversely proportional to media strength. To achieve properties of both efficiency and strength, we laminated rolling filter media with the desired properties. It is not true that the use of rolling prefilters solely to protect HEPA filters from fire-generated smoke aerosols is cost effective in every type of containment system, especially if standard fire-protection systems are available in the space. But in areas of high fire risk, where the potential fuel load is large and ignition sources are plentiful, the complication of a rolling prefilter in exit ventilation ducts to protect HEPA filters from smoke aerosols is definitely justified

  10. Electric energy supply systems: description of available technologies

    Energy Technology Data Exchange (ETDEWEB)

    Eisenhauer, J.L.; Rogers, E.A.; King, J.C.; Stegen, G.E.; Dowis, W.J.

    1985-02-01

    When comparing coal transportation with electric transmission as a means of delivering electric power, it is desirable to compare entire energy systems rather than just the transportation/transmission components because the requirements of each option may affect the requirements of other energy system components. PNL's assessment consists of two parts. The first part, which is the subject of this document, is a detailed description of the technical, cost, resource and environmental characteristics of each system component and technologies available for these components. The second part is a computer-based model that PNL has developed to simulate construction and operation of alternative system configurations and to compare the performance of these systems under a variety of economic and technical conditions. This document consists of six chapters and two appendices. A more thorough description of coal-based electric energy systems is presented in the Introduction and Chapter 1. Each of the subsequent chapters describes technologies for five system components: Western coal resources (Chapter 2), coal transportation (Chapter 3), coal gasification and gas transmission (Chapter 4), and electric power transmission (Chapter 6).

  11. CASK/MSC/WP PREPARATION SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    S. Drummond

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the Cask/MSC/WP preparation system and their bases to allow the design effort to proceed to license application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design, as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This type of SDD both leads and trails the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. This SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. This SDD addresses the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]) requirements. Additional PRD requirements may be cited, as applicable, to drive the design of specific aspects of the system, with justifications provided in the basis. Functional and operational requirements applicable to this system are obtained from the ''Project Functional and Operational Requirements'' (F and OR) (Curry 2004 [DIRS 170557]) document. Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (BSC 2004 [DIRS 171599]) and the preclosure safety analyses

  12. CASK/MSC/WP PREPARATION SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    S. Drummond

    2005-04-12

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the Cask/MSC/WP preparation system and their bases to allow the design effort to proceed to license application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design, as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This type of SDD both leads and trails the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. This SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. This SDD addresses the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]) requirements. Additional PRD requirements may be cited, as applicable, to drive the design of specific aspects of the system, with justifications provided in the basis. Functional and operational requirements applicable to this system are obtained from the ''Project Functional and Operational Requirements'' (F&OR) (Curry 2004 [DIRS 170557]) document. Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (BSC 2004 [DIRS 171599]) and the preclosure safety analyses.

  13. IDAPS (Image Data Automated Processing System) System Description

    Science.gov (United States)

    1988-06-24

    This document describes the physical configuration and components used in the image processing system referred to as IDAPS (Image Data Automated ... Processing System). This system was developed by the Environmental Research Institute of Michigan (ERIM) for Eglin Air Force Base. The system is designed

  14. Mathematical structure of ocean container transport systems; Kaiyo container yuso system no suriteki kozo ni tsuite

    Energy Technology Data Exchange (ETDEWEB)

    Shinkai, A [Kyushu University, Fukuoka (Japan). Faculty of Engineering; Chikushi, Y [Nippon Telegraph and Telephone Corp., Tokyo (Japan)

    1997-10-01

    Mathematical structure of a vessel arrangement program was discussed in order to learn roles of container ships in ocean transport systems among China, NIES/ASEAN countries and Japan. Formulation is possible on a mathematical handling method for sailing route connection diagrams between ports, a transport network to indicate container movements, a service network to indicate sailing routes, and a network generalizing them. This paper describes an analysis made on the container transport system between Japan and China, taken as an example. Four ports were selected each from Japan and China, and the statistical database for fiscals 1996 and 1994 was utilized to set models for: (a) the liner network system with transshipment at the port of Shanghai and (b) the cruising route system going through the ports of Yokohama, Nagoya and Kobe. A hypothesis was set that a consortium (coordinated ship allocation) can be implemented ideally and completely. The transport network (a) is lower by 10% in total cost than the transport network (b), resulting in 1.6 times greater productivity. Actual service network is closer to the network (b), but the system can be utilized for discussing guidelines on vessel arrangement programs with which shipping companies pursue better management efficiency under a condition that the consortium can be formed. 10 refs., 6 figs., 2 tabs.

  15. Towards unification of product and enterprise system descriptions

    CSIR Research Space (South Africa)

    Erasmus, J

    2015-06-01

    Full Text Available are utilised by enterprises and some product systems contain entire businesses, such as the operating and maintenance business of a power station. Thus, products are part of enterprises, but enterprises may also be part of product systems. To enable the design...

  16. Breeding description for fast reactors and symbiotic reactor systems

    International Nuclear Information System (INIS)

    Hanan, N.A.

    1979-01-01

    A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

  17. Process Description for the Retrieval of Earth Covered Transuranic (TRU) Waste Containers at the Hanford Site

    International Nuclear Information System (INIS)

    DEROSA, D.C.

    2000-01-01

    This document describes process and operational options for retrieval of the contact-handled suspect transuranic waste drums currently stored below grade in earth-covered trenches at the Hanford Site. Retrieval processes and options discussed include excavation, container retrieval, venting, non-destructive assay, criticality avoidance, incidental waste handling, site preparation, equipment, and shipping

  18. Process Description for the Retrieval of Earth Covered Transuranic (TRU) Waste Containers at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    DEROSA, D.C.

    2000-01-13

    This document describes process and operational options for retrieval of the contact-handled suspect transuranic waste drums currently stored below grade in earth-covered trenches at the Hanford Site. Retrieval processes and options discussed include excavation, container retrieval, venting, non-destructive assay, criticality avoidance, incidental waste handling, site preparation, equipment, and shipping.

  19. Rock Visualization System. Technical description (RVS version 3.8)

    International Nuclear Information System (INIS)

    Curtis, P.; Elfstroem, M.; Markstroem, I.

    2007-06-01

    The Rock Visualization System (RVS) has been developed by SKB for use in visualizing geological and engineering data in 3D. The purpose of this report is to provide a technical description of RVS aimed at potential program users and interested parties as well as fulfilling the function of a more general RVS reference that can be cited when writing other technical reports. The report describes RVS version 4.0. Updated versions of this report or addenda will be made available following further development of RVS and the release of subsequent versions of the program. The report covers the following main items: Technical description of the program with illustrations and examples. Limitations of the program and of functionality. For most RVS functions step-by-step tutorials are available describing how a particular function can be used to carry out a specific task. A complete set of updated tutorials is issued with each new version release of the RVS program. However, the tutorials do not cover all the possible uses of all the individual functions but rather give an overall view of their functionality. A detailed description of every RVS function and how it can be used is included in the RVS online Help system

  20. Rock Visualization System. Technical description (RVS version 3.8)

    Energy Technology Data Exchange (ETDEWEB)

    Curtis, P.; Elfstroem, M.; Markstroem, I. [FB Engineering, Goeteborg (Sweden)

    2005-04-01

    The Rock Visualization System (RVS) has been developed by SKB for use in visualizing geological and engineering data in 3D. The purpose of this report is to provide a technical description of RVS aimed at potential program users and interested parties as well as fulfilling the function of a more general RVS reference that can be cited when writing other technical reports. The report describes RVS version 3.8. Updated versions of this report or addenda will be made available following further development of RVS and the release of subsequent versions of the program. The report covers the following main items: Technical description of the program with illustrations and examples. Limitations of the program and of functionality. For most RVS functions step-by-step tutorials are available describing how a particular function can be used to carryout a specific task. A complete set of updated tutorials is issued with each new version release of the RVS program. However, the tutorials do not cover all the possible uses of all the individual functions but rather give an overall view of their functionality. A detailed description of every RVS function and how it can be used is included in the RVS online Help system.

  1. Rock Visualization System. Technical description (RVS version 3.8)

    Energy Technology Data Exchange (ETDEWEB)

    Curtis, P.; Elfstroem, M.; Markstroem, I. [Golder Associates AB (Sweden)

    2007-06-15

    The Rock Visualization System (RVS) has been developed by SKB for use in visualizing geological and engineering data in 3D. The purpose of this report is to provide a technical description of RVS aimed at potential program users and interested parties as well as fulfilling the function of a more general RVS reference that can be cited when writing other technical reports. The report describes RVS version 4.0. Updated versions of this report or addenda will be made available following further development of RVS and the release of subsequent versions of the program. The report covers the following main items: Technical description of the program with illustrations and examples. Limitations of the program and of functionality. For most RVS functions step-by-step tutorials are available describing how a particular function can be used to carry out a specific task. A complete set of updated tutorials is issued with each new version release of the RVS program. However, the tutorials do not cover all the possible uses of all the individual functions but rather give an overall view of their functionality. A detailed description of every RVS function and how it can be used is included in the RVS online Help system.

  2. Rock Visualization System. Technical description (RVS version 3.8)

    International Nuclear Information System (INIS)

    Curtis, P.; Elfstroem, M.; Markstroem, I.

    2005-04-01

    The Rock Visualization System (RVS) has been developed by SKB for use in visualizing geological and engineering data in 3D. The purpose of this report is to provide a technical description of RVS aimed at potential program users and interested parties as well as fulfilling the function of a more general RVS reference that can be cited when writing other technical reports. The report describes RVS version 3.8. Updated versions of this report or addenda will be made available following further development of RVS and the release of subsequent versions of the program. The report covers the following main items: Technical description of the program with illustrations and examples. Limitations of the program and of functionality. For most RVS functions step-by-step tutorials are available describing how a particular function can be used to carryout a specific task. A complete set of updated tutorials is issued with each new version release of the RVS program. However, the tutorials do not cover all the possible uses of all the individual functions but rather give an overall view of their functionality. A detailed description of every RVS function and how it can be used is included in the RVS online Help system

  3. PROJECT W-551 INTERIM PRETREATMENT SYSTEM PRECONCEPTUAL CANDIDATE TECHNOLOGY DESCRIPTIONS

    Energy Technology Data Exchange (ETDEWEB)

    MAY TH

    2008-08-12

    The Office of River Protection (ORP) has authorized a study to recommend and select options for interim pretreatment of tank waste and support Waste Treatment Plant (WTP) low activity waste (LAW) operations prior to startup of all the WTP facilities. The Interim Pretreatment System (IPS) is to be a moderately sized system which separates entrained solids and 137Cs from tank waste for an interim time period while WTP high level waste vitrification and pretreatment facilities are completed. This study's objective is to prepare pre-conceptual technology descriptions that expand the technical detail for selected solid and cesium separation technologies. This revision includes information on additional feed tanks.

  4. Organization model and formalized description of nuclear enterprise information system

    International Nuclear Information System (INIS)

    Yuan Feng; Song Yafeng; Li Xudong

    2012-01-01

    Organization model is one of the most important models of Nuclear Enterprise Information System (NEIS). Scientific and reasonable organization model is the prerequisite that NEIS has robustness and extendibility, and is also the foundation of the integration of heterogeneous system. Firstly, the paper describes the conceptual model of the NEIS on ontology chart, which provides a consistent semantic framework of organization. Then it discusses the relations between the concepts in detail. Finally, it gives the formalized description of the organization model of NEIS based on six-tuple array. (authors)

  5. Computer systems and software description for gas characterization system

    International Nuclear Information System (INIS)

    Vo, C.V.

    1997-01-01

    The Gas Characterization System Project was commissioned by TWRS management with funding from TWRS Safety, on December 1, 1994. The project objective is to establish an instrumentation system to measure flammable gas concentrations in the vapor space of selected watch list tanks, starting with tank AN-105 and AW-101. Data collected by this system is meant to support first tank characterization, then tank safety. System design is premised upon Characterization rather than mitigation, therefore redundancy is not required

  6. Spent nuclear fuel project cold vacuum drying facility process water conditioning system design description

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1998-01-01

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Process Water Conditioning (PWC) System. The SDD was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), the HNF-SD-SNF-DRD-O02, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the PWC equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SDD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  7. Spent nuclear fuel project cold vacuum drying facility vacuum and purge system design description

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    1998-11-30

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Vacuum and Purge System (VPS) . The SDD was developed in conjunction with HNF-SD-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), The HNF-SD-SNF-DRD-002, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the VPS equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SDD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  8. Spent nuclear fuel project cold vacuum drying facility vacuum and purge system design description

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1998-01-01

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Vacuum and Purge System (VPS) . The SDD was developed in conjunction with HNF-SD-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), The HNF-SD-SNF-DRD-002, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the VPS equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SDD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  9. Surface system Forsmark. Site descriptive modelling SDM-Site Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    Lindborg, Tobias [ed.

    2008-12-15

    % of some of the delineated sub-catchments. No major water courses flow through the central part of the site investigation area. Many brooks in the area have been deepened for considerable distances for drainage purposes. The horizontal hydraulic conductivity and specific yield of the till, the values of which are based on measurements, are typical or slightly higher than in the surrounding region. Groundwater levels in Quaternary deposits are very shallow, on average less than 0.7 m below ground during 50% of the time. Post-glacial land uplift, in combination with the flat topography, implies fast shoreline displacement. This has resulted in a young terrestrial system that contains a number of newborn lakes and wetlands. The recently isolated and shallow oligotrophic hardwater lakes that are typical for the area are unique in Sweden.The marine ecosystem at Forsmark is situated in a relatively productive coastal area in a region of otherwise fairly low primary production. The seabed is dominated by erosion and transport bottoms with heterogeneous and mobile sediments, consisting mainly of sand and gravel with varying fractions of glacial clay. Based on an overall conceptual model, it was possible to identify pools and fluxes of elements in the landscape that are of potential relevance for a safety assessment. The quantification of these elements, using both field- and model-based estimates, makes it possible to determine the relative importance of the different ecosystems with regard to elemental transport and accumulation. A special emphasis has been put on the description of transport and accumulation of organic matter, since detailed knowledge on the carbon dynamics provides a way of analysing how different ecosystem components are linked to each other through fluxes of energy, i.e. carbon. This provides a baseline for making predictions of dispersal and accumulation of matter, including radionuclides, within and between ecosystems. By this approach, the safety

  10. Plutonium Protection System (PPS). Volume 2. Hardware description. Final report

    International Nuclear Information System (INIS)

    Miyoshi, D.S.

    1979-05-01

    The Plutonium Protection System (PPS) is an integrated safeguards system developed by Sandia Laboratories for the Department of Energy, Office of Safeguards and Security. The system is designed to demonstrate and test concepts for the improved safeguarding of plutonium. Volume 2 of the PPS final report describes the hardware elements of the system. The major areas containing hardware elements are the vault, where plutonium is stored, the packaging room, where plutonium is packaged into Container Modules, the Security Operations Center, which controls movement of personnel, the Material Accountability Center, which maintains the system data base, and the Material Operations Center, which monitors the operating procedures in the system. References are made to documents in which details of the hardware items can be found

  11. Post Fukushima requirement of containment filtered venting system in NPPS

    International Nuclear Information System (INIS)

    Deo, Anuj Kumar; Bera, S.; Nagrale, D.B.; Lakshmanan, S.P.; Baburajan, P.K.; Paul, U.K.; Gaikwad, A.J.

    2015-01-01

    Post Fukushima safety enhancement through provision of an additional layer of Defence-in-Depth in the existing and new Indian nuclear power plants has led to the need of containment filtered venting system (CFVS). The regulatory review of the design of CFVS is in progress. In order to assess the same, the regulatory knowledge base had to be generated on the current state of the art of the design of such a system by study of the international experience on this system available in the open literature. The regulatory stand on requirements and implementation status of the CFVS in various countries were also studied. The information available on design features of various kinds of venting systems, relevant design basis and/or acceptance criteria were collected for supporting the design safety review of the Indian CFVS under consideration. During the on-going regulatory review process several analyses have been carried out, some more are in progress, to support the deliberations and decision making. This paper presents the above mentioned information and the summary of the analyses carried out including the status and outcome. Important aspects of the design review and associated analyses are also presented in this paper which includes the descriptions of the work on CFD study of venturi atomization, thermal hydraulics studies, shielding analysis and source term estimation studies carried out by the regulatory body. (author)

  12. Technical description of the burn-up software system MOP

    International Nuclear Information System (INIS)

    Schutte, C.K.

    1991-05-01

    The burn-up software system MOP is a research tool primary intended to study the behaviour of fission products in any reactor composition. Input data are multi-group cross-sections and data concerning the nuclide chains. An option is available to calculate a fundamental mode neutron spectrum for the specified reactor composition. A separate program can test the consistency of the specified nuclide chains. Options are available to calculate time-dependent cross-sections of lumped fission products and to take account of the leakage of gaseous fission products from the reactor core. The system is written in FORTRAN77 for a CYBER computer, using the operating system NOS/BE. The report gives a detailed technical description of the applied algorithms and the flow and storage of data. Information is provided for adapting the system to other computer configurations. (author). 5 refs.; 11 figs

  13. Non-reversible evolution of quantum chaotic system. Kinetic description

    International Nuclear Information System (INIS)

    Chotorlishvili, L.; Skrinnikov, V.

    2008-01-01

    It is well known that the appearance of non-reversibility in classical chaotic systems is connected with a local instability of phase trajectories relatively to a small change of initial conditions and parameters of the system. Classical chaotic systems reveal an exponential sensitivity to these changes. This leads to an exponential growth of initial error with time, and as the result after the statistical averaging over this error, the dynamics of the system becomes non-reversible. In spite of this, the question about the origin of non-reversibility in quantum case remains actual. The point is that the classical notion of instability of phase trajectories loses its sense during quantum consideration. The current work is dedicated to the clarification of the origin of non-reversibility in quantum chaotic systems. For this purpose we study a non-stationary dynamics of the chaotic quantum system. By analogy with classical chaos, we consider an influence of a small unavoidable error of the parameter of the system on the non-reversibility of the dynamics. It is shown in the Letter that due to the peculiarity of chaotic quantum systems, the statistical averaging over the small unavoidable error leads to the non-reversible transition from the pure state into the mixed one. The second part of the Letter is dedicated to the kinematic description of the chaotic quantum-mechanical system. Using the formalism of superoperators, a muster kinematic equation for chaotic quantum system was obtained from Liouville equation under a strict mathematical consideration

  14. Licensee Event Report system. Description of system and guidelines for reporting. Suppl. 1

    International Nuclear Information System (INIS)

    1984-02-01

    On July 26, 1983, the Commission published in the Federal Register a final rule (10 CFR 50.73) that modified and codified the Licensee Event Report (LER) system. The rule became effective on January 1, 1984. In September 1983, the NRC published NUREG-1022 which provides supporting information and guidance that is of interest to persons responsible for the preparation and review of LERs. The information contained in NUREG-1022 includes: (1) a brief description of how LERs are analyzed by the NRC, (2) a restatement of the guidance contained in the Statement of Consideration that accompanied the publication of the LER rule, (3) a set of examples of potentially reportable events with staff comments on the actual reportability of each event, (4) guidance on how to prepare an LER, including the LER forms, and (5) guidance on submittal of LERs. Subsequently, during the period from October 25, 1983 to November 16, 1983, the NRC staff held five regional meetings to discuss the scope and content of the LER rule with utility and NRC regional representatives. During these meetings numerous questions arose and were answered. This supplement to NUREG-1022 contains a summary of the questions asked and the answers given

  15. Model Based Control of Reefer Container Systems

    DEFF Research Database (Denmark)

    Sørensen, Kresten Kjær

    This thesis is concerned with the development of model based control for the Star Cool refrigerated container (reefer) with the objective of reducing energy consumption. This project has been carried out under the Danish Industrial PhD programme and has been financed by Lodam together with the Da......This thesis is concerned with the development of model based control for the Star Cool refrigerated container (reefer) with the objective of reducing energy consumption. This project has been carried out under the Danish Industrial PhD programme and has been financed by Lodam together...

  16. Simulation model for wind energy storage systems. Volume III. Program descriptions. [SIMWEST CODE

    Energy Technology Data Exchange (ETDEWEB)

    Warren, A.W.; Edsinger, R.W.; Burroughs, J.D.

    1977-08-01

    The effort developed a comprehensive computer program for the modeling of wind energy/storage systems utilizing any combination of five types of storage (pumped hydro, battery, thermal, flywheel and pneumatic). An acronym for the program is SIMWEST (Simulation Model for Wind Energy Storage). The level of detail of SIMWEST is consistent with a role of evaluating the economic feasibility as well as the general performance of wind energy systems. The software package consists of two basic programs and a library of system, environmental, and load components. Volume III, the SIMWEST program description contains program descriptions, flow charts and program listings for the SIMWEST Model Generation Program, the Simulation program, the File Maintenance program and the Printer Plotter program. Volume III generally would not be required by SIMWEST user.

  17. Description and evaluation of the Hanford personnel dosimeter program from 1944 through 1989. [Contain Glossary

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, R.H.; Fix, J.J.; Baumgartner, W.V.; Nichols, L.L.

    1990-09-01

    This report describes the evolution of personnel dosimeter technology at Hanford since the inception of Hanford operations in 1944. Each of the personnel dosimeter systems used by people working or visiting Hanford is described. In addition, the procedures used to calibrate and calculate dose for each of the dosimeter systems are described. The accuracy of the recorded dose, primarily whole body deep dose, for the different dosimeter systems is evaluated. The evaluation is based on an extensive review of historical literature, as well as a 1989 intercomparison study of all film dosimeters and performance testing of the thermoluminescent dosimeter, also conducted during 1989. 73 refs., 40 figs., 41 tabs.

  18. Description of Transmutation Library for Fuel Cycle System Analyses

    International Nuclear Information System (INIS)

    Piet, Steven J.; Bays, Samuel E.; Hoffman, Edward A.

    2010-01-01

    This report documents the Transmutation Library that is used in Fuel Cycle System Analyses. This version replaces the 2008 version.(Piet2008) The Transmutation Library has the following objectives: (1) Assemble past and future transmutation cases for system analyses. (2) For each case, assemble descriptive information such as where the case was documented, the purpose of the calculation, the codes used, source of feed material, transmutation parameters, and the name of files that contain raw or source data. (3) Group chemical elements so that masses in separation and waste processes as calculated in dynamic simulations or spreadsheets reflect current thinking of those processes. For example, the CsSr waste form option actually includes all Group 1A and 2A elements. (4) Provide mass fractions at input (charge) and output (discharge) for each case. (5) Eliminate the need for either ''fission product other'' or ''actinide other'' while conserving mass. Assessments of waste and separation cannot use ''fission product other'' or ''actinide other'' as their chemical behavior is undefined. (6) Catalog other isotope-specific information in one place, e.g., heat and dose conversion factors for individual isotopes. (7) Describe the correlations for how input and output compositions change as a function of UOX burnup (for LWR UOX fuel) or fast reactor (FR) transuranic (TRU) conversion ratio (CR) for either FR-metal or FR-oxide. This document therefore includes the following sections: (1) Explanation of the data set information, i.e., the data that describes each case. In no case are all of the data presented in the Library included in previous documents. In assembling the Library, we return to raw data files to extract the case and isotopic data, into the specified format. (2) Explanation of which isotopes and elements are tracked. For example, the transition metals are tracked via the following: two Zr isotopes, Zr-other, Tc99, Tc-other, two Mo-Ru-Rh-Pd isotopes, Mo

  19. On the validity of collective variable description of Bose systems

    International Nuclear Information System (INIS)

    Takahashi, Minoru

    1975-01-01

    The validity of Sunakawa, Yamasaki and Kebukawa's Hamiltonian and that of Bogoliubov and Zubarev's Hamiltonian are examined. Perturbational expansion of the ground state energy by these Hamiltonians disagrees with the exact solution of Lieb and Liniger for one-dimensional Bose system with repulsive delta-function interaction. This fact suggests that these Hamiltonians are not microscopic descriptions of the many-Boson system. Mathematical inconsistency in Bogoliubov and Zubarev's theory is also pointed out. Moreover analytic expression of high density expansion for the ground state energy density e 0 is found out to be e 0 n -3 =γ-(4/3π)γsup(3/2)+(1/6-1/π 2 )γ 2 +O(γsup(5/2)), γ=c/n, for one-dimensional Bose system with delta function interaction (density n, strength 2c, h=2m=1) by the use of the correlated basis function method. (auth.)

  20. Breakdown of the classical description of a local system

    DEFF Research Database (Denmark)

    Eran, Kot; Grønbech-Jensen, Niels; Nielsen, Bo Melholt

    2012-01-01

    We provide a straightforward demonstration of a fundamental difference between classical and quantum mechanics for a single local system: namely, the absence of a joint probability distribution of the position x and momentum p. Elaborating on a recently reported criterion by Bednorz and Belzig...... of the breakdown of a classical description of the underlying state. Most importantly, the criterion used does not rely on quantum mechanics and can thus be used to demonstrate nonclassicality of systems not immediately apparent to exhibit quantum behavior. The criterion is directly applicable to any system...... [ Phys. Rev. A 83 052113 (2011)] we derive a simple criterion that must be fulfilled for any joint probability distribution in classical physics. We demonstrate the violation of this criterion using the homodyne measurement of a single photon state, thus proving a straightforward signature...

  1. SWEPP Gamma-Ray Spectrometer System software design description

    International Nuclear Information System (INIS)

    Femec, D.A.; Killian, E.W.

    1994-08-01

    To assist in the characterization of the radiological contents of contract-handled waste containers at the Stored Waste Examination Pilot Plant (SWEPP), the SWEPP Gamma-Ray Spectrometer (SGRS) System has been developed by the Radiation Measurements and Development Unit of the Idaho National Engineering Laboratory. The SGRS system software controls turntable and detector system activities. In addition to determining the concentrations of gamma-ray-emitting radionuclides, this software also calculates attenuation-corrected isotopic mass ratios of-specific interest. This document describes the software design for the data acquisition and analysis software associated with the SGRS system

  2. SWEPP Gamma-Ray Spectrometer System software design description

    Energy Technology Data Exchange (ETDEWEB)

    Femec, D.A.; Killian, E.W.

    1994-08-01

    To assist in the characterization of the radiological contents of contract-handled waste containers at the Stored Waste Examination Pilot Plant (SWEPP), the SWEPP Gamma-Ray Spectrometer (SGRS) System has been developed by the Radiation Measurements and Development Unit of the Idaho National Engineering Laboratory. The SGRS system software controls turntable and detector system activities. In addition to determining the concentrations of gamma-ray-emitting radionuclides, this software also calculates attenuation-corrected isotopic mass ratios of-specific interest. This document describes the software design for the data acquisition and analysis software associated with the SGRS system.

  3. Cold Vacuum Drying facility crane and hoist system design description

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) crane and hoist system. The overhead crane and hoist system is located in the process bays of the CVDF. It supports the processes required to drain the water and dry the spent nuclear fuel contained in the multi-canister overpacks after they have been removed from the K-Basins. The cranes will also be used to assist maintenance activities within the bays, as required

  4. Systems for the monitoring of working conditions relating to health and safety : extensive descriptions : Belgium, Germany, Luxembourg, The Netherlands

    NARCIS (Netherlands)

    Prins, R.; Verboon, F.

    1991-01-01

    This report contains the extensive descriptions of (some of) the monitoring systems on health and safety in use in the Benelux countries and Germany. The project of which this report is a part aimed at gathering information on monitoring systems throughout the EC member states.

  5. Containment systems for uranium-mill tailings

    International Nuclear Information System (INIS)

    Hartley, J.N.; Buelt, J.L.

    1982-11-01

    Cover and liner systems for uranium mill tailings in the United States must satisfy stringent requirements regarding long-term stability, radon control, and radionuclide and hazardous chemical migration. The cover and liner technology discussed in this paper involves: (1) single and multilayer earthen cover systems; (2) asphalt emulsion radon barrier systems; and (3) asphalt, clay, and synthetic liner systems. These systems have been field tested at the Grand Junction, Colorado, tailings pile, where they have been shown to effectively reduce radon releases and radionuclide and chemical migration

  6. A brief description of Polanco`s hybrid system

    Energy Technology Data Exchange (ETDEWEB)

    Nunes, Ventura [Instituto de Ingenieria Electrica, Universidad de la Republica, Montevideo (Uruguay)

    1997-12-31

    Since 1995, a hybrid system wind - PV is in service in Polanco, Uruguay. A brief description of this system and the criteria employed in its design are outlined. The experience obtained during two years operation are described from the points of view of the equipment reliability and of the electrical service provided. Future prospects of this kind of installations in Uruguay as part of the rural electrification policy are presented. [Espanol] Desde 1995 esta en servicio en Polanco, Uruguay un sistema hibrido viento-fotovoltaico. Aqui se describe brevemente este sistema y el criterio empleado en su diseno. Se describe la experiencia obtenida durante dos anos de operacion desde el punto de vista de confiabilidad de los equipos y del servicio electrico proporcionado. Se presentan prospectos futuros de esta clase de instalaciones en Uruguay, como parte de la politica de electrificacion rural.

  7. A brief description of Polanco`s hybrid system

    Energy Technology Data Exchange (ETDEWEB)

    Nunes, Ventura [Instituto de Ingenieria Electrica, Universidad de la Republica, Montevideo (Uruguay)

    1998-12-31

    Since 1995, a hybrid system wind - PV is in service in Polanco, Uruguay. A brief description of this system and the criteria employed in its design are outlined. The experience obtained during two years operation are described from the points of view of the equipment reliability and of the electrical service provided. Future prospects of this kind of installations in Uruguay as part of the rural electrification policy are presented. [Espanol] Desde 1995 esta en servicio en Polanco, Uruguay un sistema hibrido viento-fotovoltaico. Aqui se describe brevemente este sistema y el criterio empleado en su diseno. Se describe la experiencia obtenida durante dos anos de operacion desde el punto de vista de confiabilidad de los equipos y del servicio electrico proporcionado. Se presentan prospectos futuros de esta clase de instalaciones en Uruguay, como parte de la politica de electrificacion rural.

  8. Description of a 20 Kilohertz power distribution system

    Science.gov (United States)

    Hansen, I. G.

    1986-01-01

    A single phase, 440 VRMS, 20 kHz power distribution system with a regulated sinusoidal wave form is discussed. A single phase power system minimizes the wiring, sensing, and control complexities required in a multi-sourced redundantly distributed power system. The single phase addresses only the distribution link; mulitphase lower frequency inputs and outputs accommodation techniques are described. While the 440 V operating potential was initially selected for aircraft operating below 50,000 ft, this potential also appears suitable for space power systems. This voltage choice recognizes a reasonable upper limit for semiconductor ratings, yet will direct synthesis of 220 V, 3 power. A 20 kHz operating frequency was selected to be above the range of audibility, minimize the weight of reactive components, yet allow the construction of single power stages of 25 to 30 kW. The regulated sinusoidal distribution system has several advantages. With a regulated voltage, most ac/dc conversions involve rather simple transformer rectifier applications. A sinusoidal distribution system, when used in conjunction with zero crossing switching, represents a minimal source of EMI. The present state of 20 kHz power technology includes computer controls of voltage and/or frequency, low inductance cable, current limiting circuit protection, bi-directional power flow, and motor/generator operating using standard induction machines. A status update and description of each of these items and their significance is presented.

  9. The application of PLC in 60Co container inspection system

    International Nuclear Information System (INIS)

    Huang Yibin; Xiang Xincheng

    2001-01-01

    The author discusses the interlock technique of 60 Co container inspection system, and introduces the hardware structure and program of interlock control system using PLC. Due to adopting PLC distributed control, the system works stably and reliably. The successful application of PLC in 60 Co container inspection system has some use for reference in nuclear technology field

  10. 46 CFR 56.50-60 - Systems containing oil.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Systems containing oil. 56.50-60 Section 56.50-60... APPURTENANCES Design Requirements Pertaining to Specific Systems § 56.50-60 Systems containing oil. (a)(1) Oil-piping systems for the transfer or discharge of cargo or fuel oil must be separate from other piping...

  11. Container corporation: The financial, management, and operating implications of a universal beverage container recovery system: Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    1991-01-01

    This report contains an overview of the system. It discusses containers and container materials, elements of the system, and the container recovery cost structure. It includes a financial evaluation and addresses issues of implementation.

  12. Leakage detection system in nuclear reactor container

    International Nuclear Information System (INIS)

    Kurosawa, Masahiko.

    1993-01-01

    The present invention comprises an injection means for adding radioactive materials to coolants in a container cooler, a gamma ray amplitude analyzer connected to coolant pipelines and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to a drain water sump and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to various kinds of pipelines and a means for recording/transferring the data of the result of the measurement, and a data processing means for comparing and analyzing the measured data of each of the gamma ray amplitude analyzers inputted from each of date recording/transferring means. The gamma ray amplitude analysis for each of the pipelines and drain water sump are conducted at an appropriate frequency, and the measured data are compared and analyzed, to improve the detection accuracy for a trace amount of leakage from each of the pressure pipelines and the container cooler coolant pipelines, thereby enabling to specify the pipeline having leakage. Maintenance efficiency is improved, and severe rupture accident in each of pressure pipelines is prevented previously. (N.H.)

  13. Interrelations between different canonical descriptions of dissipative systems

    International Nuclear Information System (INIS)

    Schuch, D; Guerrero, J; López-Ruiz, F F; Aldaya, V

    2015-01-01

    There are many approaches for the description of dissipative systems coupled to some kind of environment. This environment can be described in different ways; only effective models are being considered here. In the Bateman model, the environment is represented by one additional degree of freedom and the corresponding momentum. In two other canonical approaches, no environmental degree of freedom appears explicitly, but the canonical variables are connected with the physical ones via non-canonical transformations. The link between the Bateman approach and those without additional variables is achieved via comparison with a canonical approach using expanding coordinates, as, in this case, both Hamiltonians are constants of motion. This leads to constraints that allow for the elimination of the additional degree of freedom in the Bateman approach. These constraints are not unique. Several choices are studied explicitly, and the consequences for the physical interpretation of the additional variable in the Bateman model are discussed. (paper)

  14. Surface energy and radiation balance systems - General description and improvements

    Science.gov (United States)

    Fritschen, Leo J.; Simpson, James R.

    1989-01-01

    Surface evaluation of sensible and latent heat flux densities and the components of the radiation balance were desired for various vegetative surfaces during the ASCOT84 experiment to compare with modeled results and to relate these values to drainage winds. Five battery operated data systems equipped with sensors to determine the above values were operated for 105 station days during the ASCOT84 experiment. The Bowen ratio energy balance technique was used to partition the available energy into the sensible and latent heat flux densities. A description of the sensors and battery operated equipment used to collect and process the data is presented. In addition, improvements and modifications made since the 1984 experiment are given. Details of calculations of soil heat flow at the surface and an alternate method to calculate sensible and latent heat flux densities are provided.

  15. Interrelations between different canonical descriptions of dissipative systems

    Science.gov (United States)

    Schuch, D.; Guerrero, J.; López-Ruiz, F. F.; Aldaya, V.

    2015-04-01

    There are many approaches for the description of dissipative systems coupled to some kind of environment. This environment can be described in different ways; only effective models are being considered here. In the Bateman model, the environment is represented by one additional degree of freedom and the corresponding momentum. In two other canonical approaches, no environmental degree of freedom appears explicitly, but the canonical variables are connected with the physical ones via non-canonical transformations. The link between the Bateman approach and those without additional variables is achieved via comparison with a canonical approach using expanding coordinates, as, in this case, both Hamiltonians are constants of motion. This leads to constraints that allow for the elimination of the additional degree of freedom in the Bateman approach. These constraints are not unique. Several choices are studied explicitly, and the consequences for the physical interpretation of the additional variable in the Bateman model are discussed.

  16. Development on design methodology of PWR passive containment system

    International Nuclear Information System (INIS)

    Lee, Seong Wook

    1998-02-01

    The containment is the most important barrier against the release of radioactive materials into the environment during accident conditions of nuclear power plants. Therefore the development of a reliable containment cooling system is one of key areas in advanced reactor development. To enhance the safety of the containment system, many new containment system designs have been proposed and developed in the world. Several passive containment cooling system (PCCS) concepts for both steel and concrete containment systems are overviewed and assessed comparatively. Major concepts considered are: (a) the spray of water on the outer surface of a steel containment from an elevated tank, (b) an external moat for a steel containment, (c) a suppression pool for a concrete containment, and (d) combination of the internal spray and internal or external condensers for a concrete containment. Emphasis is given to the heat removal principles, the required heat transfer area, system complexity and operational reliability. As one of conceptual design steps of containment, a methodology based on scaling principles is proposed to determine the containment size according to the power level. The AP600 containment system is selected as the reference containment to which the scaling laws are applied. Governing equations of containment pressure are set up in consideration of containment behavior in accident conditions. Then, the dimensionless numbers, which characterize the containment phenomena, are derived for the blowdown dominant and decay heat dominant stage, respectively. The important phenomena in blowdown stage are mass and energy sources and their absorption in containment atmosphere or containment structure, while heat transfer to the outer environment becomes important in decay heat stage. Based on their similarity between the prototype and the model, the containment sizes are determined for higher power levels and are compared with the SPWR containment design values available

  17. Dynamic testing of MFTF containment-vessel structural system

    International Nuclear Information System (INIS)

    Weaver, H.J.; McCallen, D.B.; Eli, M.W.

    1982-01-01

    Dynamic (modal) testing was performed on the Magnetic Fusion Test Facility (MFTF) containment vessel. The seismic design of this vessel was heavily dependent upon the value of structural damping used in the analysis. Typically for welded steel vessels, a value of 2 to 3% of critical is used. However, due to the large mass of the vessel and magnet supported inside, we felt that the interaction between the structure and its foundation would be enhanced. This would result in a larger value of damping because vibrational energy in the structure would be transferred through the foundation into the surrounding soil. The dynamic test performed on this structure (with the magnet in place) confirmed this later theory and resulted in damping values of approximately 4 to 5% for the whole body modes. This report presents a brief description of dynamic testing emphasizing the specific test procedure used on the MFTF-A system. It also presents an interpretation of the damping mechanisms observed (material and geometric) based upon the spatial characteristics of the modal parameters

  18. CLASSIFICATION OF THE MGR NON-FUEL COMPONENTS DISPOSAL CONTAINER SYSTEM

    International Nuclear Information System (INIS)

    J.A. Ziegler

    1999-01-01

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

  19. Contained fission explosion breeder reactor system

    International Nuclear Information System (INIS)

    Juhl, N.H.; Marwick, E.F.

    1983-01-01

    A reactor system for producing useful thermal energy and valuable isotopes, such as plutonium-239, uranium-233, and/or tritium, in which a pair of sub-critical masses of fissile and fertile actinide slugs are propelled into an ellipsoidal pressure vessel. The propelled slugs intercept near the center of the chamber where the concurring slugs become a more than prompt configuration thereby producing a fission explosion. Re-useable accelerating mechanisms are provided external of the vessel for propelling the slugs at predetermined time intervals into the vessel. A working fluid of lean molten metal slurry is injected into the chamber prior to each explosion for the attenuation of the explosion's effects, for the protection of the chamber's walls, and for the absorbtion of thermal energy and debris from the explosion. The working fluid is injected into the chamber in a pattern so as not to interfere with the flight paths of the slugs and to maximize the concentration of working fluid near the chamber's center. The heated working fluid is drained from the vessel and is used to perform useful work. Most of the debris from the explosion is collected as precipitate and is used for the manufacture of new slugs

  20. Retrievable surface storage facility conceptual system design description

    Energy Technology Data Exchange (ETDEWEB)

    1977-03-01

    The studies evaluated several potentially attractive methods for processing and retrievably storing high-level radioactive waste after delivery to the Federal repository. These studies indicated that several systems could be engineered to safely store the waste, but that the simplest and most attractive concept from a technical standpoint would be to store the waste in a sealed stainless steel canister enclosed in a 2 in. thick carbon steel cask which in turn would be inserted into a reinforced concrete gamma-neutron shield, which would also provide the necessary air-cooling through an air annulus between the cask and the shield. This concept best satisfies the requirements for safety, long-term exposure to natural phenomena, low capital and operating costs, retrievability, amenability to incremental development, and acceptably small environmental impact. This document assumes that the reference site would be on ERDA's Hanford reservation. This document is a Conceptual System Design Description of the facilities which could satisfy all of the functional requirements within the established basic design criteria. The Retrievable Surface Storage Facility (RSSF) is planned with the capacity to process and store the waste received in either a calcine or glass/ceramic form. The RSSF planning is based on a modular development program in which the modular increments are constructed at rates matching projected waste receipts.

  1. Data integration, systems approach and multilevel description of complex biosystems

    International Nuclear Information System (INIS)

    Hernández-Lemus, Enrique

    2013-01-01

    Recent years have witnessed the development of new quantitative approaches and theoretical tenets in the biological sciences. The advent of high throughput experiments in genomics, proteomics and electrophysiology (to cite just a few examples) have provided the researchers with unprecedented amounts of data to be analyzed. Large datasets, however can not provide the means to achieve a complete understanding of the underlying biological phenomena, unless they are supplied with a solid theoretical framework and with proper analytical tools. It is now widely accepted that by using and extending some of the paradigmatic principles of what has been called complex systems theory, some degree of advance in this direction can be attained. We will be presenting ways in which by using data integration techniques (linear, non-linear, combinatorial, graphical), multidimensional-multilevel descriptions (multifractal modeling, dimensionality reduction, computational learning), as well as an approach based in systems theory (interaction maps, probabilistic graphical models, non-equilibrium physics) have allowed us to better understand some problems in the interface of Statistical Physics and Computational Biology

  2. Retrievable surface storage facility conceptual system design description

    International Nuclear Information System (INIS)

    1977-03-01

    The studies evaluated several potentially attractive methods for processing and retrievably storing high-level radioactive waste after delivery to the Federal repository. These studies indicated that several systems could be engineered to safely store the waste, but that the simplest and most attractive concept from a technical standpoint would be to store the waste in a sealed stainless steel canister enclosed in a 2 in. thick carbon steel cask which in turn would be inserted into a reinforced concrete gamma-neutron shield, which would also provide the necessary air-cooling through an air annulus between the cask and the shield. This concept best satisfies the requirements for safety, long-term exposure to natural phenomena, low capital and operating costs, retrievability, amenability to incremental development, and acceptably small environmental impact. This document assumes that the reference site would be on ERDA's Hanford reservation. This document is a Conceptual System Design Description of the facilities which could satisfy all of the functional requirements within the established basic design criteria. The Retrievable Surface Storage Facility (RSSF) is planned with the capacity to process and store the waste received in either a calcine or glass/ceramic form. The RSSF planning is based on a modular development program in which the modular increments are constructed at rates matching projected waste receipts

  3. LCA comparison of container systems in municipal solid waste management

    International Nuclear Information System (INIS)

    Rives, Jesus; Rieradevall, Joan; Gabarrell, Xavier

    2010-01-01

    The planning and design of integrated municipal solid waste management (MSWM) systems requires accurate environmental impact evaluation of the systems and their components. This research assessed, quantified and compared the environmental impact of the first stage of the most used MSW container systems. The comparison was based on factors such as the volume of the containers, from small bins of 60-80 l to containers of 2400 l, and on the manufactured materials, steel and high-density polyethylene (HDPE). Also, some parameters such as frequency of collections, waste generation, filling percentage and waste container contents, were established to obtain comparable systems. The methodological framework of the analysis was the life cycle assessment (LCA), and the impact assessment method was based on CML 2 baseline 2000. Results indicated that, for the same volume, the collection systems that use HDPE waste containers had more of an impact than those using steel waste containers, in terms of abiotic depletion, global warming, ozone layer depletion, acidification, eutrophication, photochemical oxidation, human toxicity and terrestrial ecotoxicity. Besides, the collection systems using small HDPE bins (60 l or 80 l) had most impact while systems using big steel containers (2400 l) had less impact. Subsequent sensitivity analysis about the parameters established demonstrated that they could change the ultimate environmental impact of each waste container collection system, but that the comparative relationship between systems was similar.

  4. Solar Pilot Plant, Phase I. Preliminary design report. Volume II. System description and system analysis. CDRL item 2

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-05-01

    Honeywell conducted a parametric analysis of the 10-MW(e) solar pilot plant requirements and expected performance and established an optimum system design. The main analytical simulation tools were the optical (ray trace) and the dynamic simulation models. These are described in detail in Books 2 and 3 of this volume under separate cover. In making design decisions, available performance and cost data were used to provide a design reflecting the overall requirements and economics of a commercial-scale plant. This volume contains a description of this analysis/design process and resultant system/subsystem design and performance.

  5. Descriptive review of tuberculosis surveillance systems across the circumpolar regions

    Directory of Open Access Journals (Sweden)

    Annie-Claude Bourgeois

    2016-04-01

    baseline knowledge on similarities and differences among circumpolar tuberculosis surveillance systems. The similarity in case definitions will allow for description of the epidemiology of TB based on surveillance data in circumpolar regions, further study of tuberculosis trends across regions, and recommendation of best practices to improve surveillance activities.

  6. 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

  7. 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

  8. Plutonium active operation of the Winfrith modular containment system

    International Nuclear Information System (INIS)

    Sanders, M.J.; Pengelly, M.G.A.; McSherry, K.

    1985-01-01

    Three gloveboxes contaminated with plutonium have been dismantled inside the Winfrith Modular Containment System. This system is a portable, demountable pressurised suit area with its own filters and shower entry tunnel. Details of the operation are given. (U.K.)

  9. An advanced dispatching technology for large container inspection system

    International Nuclear Information System (INIS)

    Chen Zhiqiang; Zhang Li; Kang Kejun; Gao Wenhuan

    2001-01-01

    The author describes the transmitting and dispatching technology of large container inspection system. It introduces the structure of the double buffer graded pipe lining used in the system. Strategies of queue mechanism and waiting dispatch policy are illustrated

  10. Report on container technology for the ATLAS TDAQ system

    CERN Document Server

    Gadirov, Hamid

    2016-01-01

    My summer student project "Container technology for the Upgrade of the ATLAS Trigger and Data Acquisition (TDAQ) system" focused on the research of container-based (operating system-level) virtualization for TDAQ software. Several tests were performed on Docker platform, all of them showed compatibility for TDAQ software.

  11. Throughput maximization of parcel sorter systems by scheduling inbound containers

    NARCIS (Netherlands)

    Haneyah, S.W.A.; Schutten, Johannes M.J.; Fikse, K.; Clausen, Uwe; ten Hompel, Michael; Meier, J. Fabian

    2013-01-01

    This paper addresses the inbound container scheduling problem for automated sorter systems in express parcel sorting. The purpose is to analyze which container scheduling approaches maximize the throughput of sorter systems. We build on existing literature, particularly on the dynamic load balancing

  12. Cold Vacuum Drying Facility Condensate Collection System Design Description. System 19

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    2000-01-01

    The Cold Vacuum Drying (CVD) Facility of Spent Nuclear Fuel (SNF) provides required process systems, supporting equipment, and facilities to support the SNF Project mission. This system design description (SDD) addresses the Condensate Collection System (CCS). This is a general service system. The CCS begins at the condensate outlet of the general process air-handling unit (AHU) and the condensate outlets for the active process bays AHUs. The system terminates at each condensate collection tank (5 total)

  13. Light and heavy water replacing system in reactor container

    International Nuclear Information System (INIS)

    Miyamoto, Keiji.

    1979-01-01

    Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

  14. Gross Containment Leakage Monitoring System (GCLM) applied to accidental impairment of containment integrity determination

    International Nuclear Information System (INIS)

    Dinu, Camelia; Talpalariu, A.; Constantinescu, G.

    2007-01-01

    The Prioritization of Generic Safety Issues (NUREG-0933 of October 2006), section 1 task II.E.4 item II.E.4.3 recommends that a method of periodic or continuous testing has to be available, in order to detect unknown gross openings in the nuclear power plants containment structure. The Palisades incident and three other incidents are exemplified, when the reactor was operated for about 1.5 years, while the containment isolation valves in a purge system bypass line were unknowingly locked in the open position. It was estimated that the presence of a GCLM system could identify an unknown breach and reduce the expected unavailability of containment due to containment integrity breach events, to a 1.6x10 -3 /year demand. (authors)

  15. Container code recognition in information auto collection system of container inspection

    International Nuclear Information System (INIS)

    Su Jianping; Chen Zhiqiang; Zhang Li; Gao Wenhuan; Kang Kejun

    2003-01-01

    Now custom needs electrical application and automatic detection. Container inspection should not only give the image of the goods, but also auto-attain container's code and weight. Its function and track control, information transfer make up the Information Auto Collection system of Container Inspection. Code Recognition is the point. The article is based on model match, the close property of character, and uses it to recognize. Base on checkout rule, design the adjustment arithmetic, form the whole recognition strategy. This strategy can achieve high recognition ratio and robust property

  16. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  17. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  18. Cold Vacuum Dryer (CVD) Facility Security System Design Description. System 54

    International Nuclear Information System (INIS)

    WHITEHURST, R.

    2000-01-01

    This system design description (SDD) addresses the Cold Vacuum Drying (CVD) Facility security system. The system's primary purpose is to provide reasonable assurance that breaches of security boundaries are detected and assessment information is provided to protective force personnel. In addition, the system is utilized by Operations to support reduced personnel radiation goals and to provide reasonable assurance that only authorized personnel are allowed to enter designated security areas

  19. Description of the moderator systems for the ESS project

    International Nuclear Information System (INIS)

    Stendal, K.

    1996-10-01

    This paper describes a suggestion for the arrangement of the Cold Neutron Sources for the two targets in the ESS project. (European Spallation Source). The suggestion is based upon the technique of the existing cold neutron sources at Risoe in Denmark, HMI and Geestacht in Germany. As moderating media all of them use H 2 in supercritical condition, circulated by blowers, and the safety of the systems is based upon the triple-containment philosophy. This seems to be the most convenient principle to use near the ESS targets, as it provides a larger degree of freedom with respect to the arrangement of these sources and pipe connections to the chambers, especially because the space is limited and access to the target is relatively complicated. The moderator chambers have been designed by KFA, Juelich and the rest of the arrangement by Risoe, DK. The price calculations used for the ESS project are based upon this arrangement. (au)

  20. Description of the moderator systems for the ESS project

    Energy Technology Data Exchange (ETDEWEB)

    Stendal, K.

    1996-10-01

    This paper describes a suggestion for the arrangement of the Cold Neutron Sources for the two targets in the ESS project. (European Spallation Source). The suggestion is based upon the technique of the existing cold neutron sources at Risoe in Denmark, HMI and Geestacht in Germany. As moderating media all of them use H{sub 2} in supercritical condition, circulated by blowers, and the safety of the systems is based upon the triple-containment philosophy. This seems to be the most convenient principle to use near the ESS targets, as it provides a larger degree of freedom with respect to the arrangement of these sources and pipe connections to the chambers, especially because the space is limited and access to the target is relatively complicated. The moderator chambers have been designed by KFA, Juelich and the rest of the arrangement by Risoe, DK. The price calculations used for the ESS project are based upon this arrangement. (au).

  1. The LHCb RICH system; detector description and operation

    Energy Technology Data Exchange (ETDEWEB)

    Papanestis, A., E-mail: antonis.papanestis@stfc.ac.uk

    2014-12-01

    Two RICH detectors provide positive charged hadron identification in the LHCb experiment at the Large Hadron Collider at CERN. RICH 1 covers the full acceptance of the spectrometer and contains two radiators: aerogel and C{sub 4}F{sub 10}. RICH 2 covers half the acceptance and uses CF{sub 4} as a Cherenkov radiator. Photon detection is performed by the Hybrid Photon Detectors (HPDs), with silicon pixel sensors and bump-bonded readout encapsulated in a vacuum tube for efficient, low-noise single photon detection. The LHCb RICH detectors form a complex system of three radiators, 120 mirrors and 484 photon detectors operating in the very challenging environment of the LHC. The high performance of the system in pion and kaon identification in the momentum range of 2–100 GeV/c is reached only after careful calibration of many parameters. Operational efficiency above 99% was achieved by a high level of automatization in the operation of the detectors, from switching-on to error recovery. The challenges of calibrating and operating such a system will be presented. - Highlights: • This paper describes the operation and calibration of the LHCb RICH detectors. • The scintillation of CF{sub 4} was successfully suppressed with CO{sub 2}. • The refractive index of the gas radiators was calibrated with data to an accuracy better than 0.1%. • The Hybrid Photons Detectors were calibrated for operation in a magnetic field without loss of resolution.

  2. Expert system for controlling plant growth in a contained environment

    Science.gov (United States)

    May, George A. (Inventor); Lanoue, Mark Allen (Inventor); Bethel, Matthew (Inventor); Ryan, Robert E. (Inventor)

    2011-01-01

    In a system for optimizing crop growth, vegetation is cultivated in a contained environment, such as a greenhouse, an underground cavern or other enclosed space. Imaging equipment is positioned within or about the contained environment, to acquire spatially distributed crop growth information, and environmental sensors are provided to acquire data regarding multiple environmental conditions that can affect crop development. Illumination within the contained environment, and the addition of essential nutrients and chemicals are in turn controlled in response to data acquired by the imaging apparatus and environmental sensors, by an "expert system" which is trained to analyze and evaluate crop conditions. The expert system controls the spatial and temporal lighting pattern within the contained area, and the timing and allocation of nutrients and chemicals to achieve optimized crop development. A user can access the "expert system" remotely, to assess activity within the growth chamber, and can override the "expert system".

  3. COMMIX analysis of AP-600 Passive Containment Cooling System

    International Nuclear Information System (INIS)

    Chang, J.F.C.; Chien, T.H.; Ding, J.; Sun, J.G.; Sha, W.T.

    1992-01-01

    COMMIX modeling and basic concepts that relate components, i.e., containment, water film cooling, and natural draft air flow systems. of the AP-600 Passive Containment Cooling System are discussed. The critical safety issues during a postulated accident have been identified as (1) maintaining the liquid film outside the steel containment vessel, (2) ensuring the natural convection in the air annulus. and (3) quantifying both heat and mass transfer accurately for the system. The lack of appropriate heat and mass transfer models in the present analysis is addressed. and additional assessment and validation of the proposed models is proposed

  4. Licensing systems and inspection of nuclear installations in NEA Member countries. Part 1, Description of licensing systems

    International Nuclear Information System (INIS)

    1977-01-01

    This study provides an assessment of the legislative and regulatory provisions applicable and of the practices followed in the countries concerned and is divided into two separate sections. This document is the first part only. It contains the description of national licensing and inspection systems for nuclear installations in the twenty OECD countries which have specific regulations in this field. Each analysis has been presented following a plan which is as standardised as possible so as to facilitate comparison between the national systems. Part II, which is not included in this document, contains the diagrams illustrating the steps in the licensing procedure and the duties of the bodies involved as well as certain additional documents. It also includes a table showing the sequence of the main steps in the licensing process in the countries covered by this Study

  5. Mass extraction container closure integrity physical testing method development for parenteral container closure systems.

    Science.gov (United States)

    Yoon, Seung-Yil; Sagi, Hemi; Goldhammer, Craig; Li, Lei

    2012-01-01

    Container closure integrity (CCI) is a critical factor to ensure that product sterility is maintained over its entire shelf life. Assuring the CCI during container closure (C/C) system qualification, routine manufacturing and stability is important. FDA guidance also encourages industry to develop a CCI physical testing method in lieu of sterility testing in a stability program. A mass extraction system has been developed to check CCI for a variety of container closure systems such as vials, syringes, and cartridges. Various types of defects (e.g., glass micropipette, laser drill, wire) were created and used to demonstrate a detection limit. Leakage, detected as mass flow in this study, changes as a function of defect length and diameter. Therefore, the morphology of defects has been examined in detail with fluid theories. This study demonstrated that a mass extraction system was able to distinguish between intact samples and samples with 2 μm defects reliably when the defect was exposed to air, water, placebo, or drug product (3 mg/mL concentration) solution. Also, it has been verified that the method was robust, and capable of determining the acceptance limit using 3σ for syringes and 6σ for vials. Sterile products must maintain their sterility over their entire shelf life. Container closure systems such as those found in syringes and vials provide a seal between rubber and glass containers. This seal must be ensured to maintain product sterility. A mass extraction system has been developed to check container closure integrity for a variety of container closure systems such as vials, syringes, and cartridges. In order to demonstrate the method's capability, various types of defects (e.g., glass micropipette, laser drill, wire) were created in syringes and vials and were tested. This study demonstrated that a mass extraction system was able to distinguish between intact samples and samples with 2 μm defects reliably when the defect was exposed to air, water

  6. System Description for the KW Basin Integrated Water Treatment System (IWTS) (70.3)

    International Nuclear Information System (INIS)

    DERUSSEAU, R.R.

    2000-01-01

    This is a description of the system that collects and processes the sludge and radioactive ions released by the spent nuclear fuel (SNF) processing operations conducted in the 105 KW Basin. The system screens, settles, filters, and conditions the basin water for reuse. Sludge and most radioactive ions are removed before the water is distributed back to the basin pool. This system is part of the Spent Nuclear Fuel Project (SNFP)

  7. Requirements for containment system components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term 'components' includes non registered items

  8. Requirements for containment system components in CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term `components` includes non registered items.

  9. Introduction to the controlled nuclear fusion (magnetic containment systems)

    International Nuclear Information System (INIS)

    Cabrera, J.A.; Guasp, J.; Martin, R.

    1975-01-01

    The magnetic containment systems, their more important features, and their potentiality to became thermonuclear reactors is described. The work is based upon the first part of a set of lectures dedicated to Plasma and Fusion Physics. (author)

  10. Application of system-process-goal approach for description of TRIGA RC-1 system

    International Nuclear Information System (INIS)

    Gadomski, Adam M.

    1986-01-01

    The new methodology of the goal oriented description of an artificial system is presented. In the SPG approach (System-Process-Goal) it is assumed that the knowledge necessary for achieving the goal is available but it is not ordered or ordered for other purposes. The aim of SPG is to give the description of the analyzed system in form of network by decomposition of goal-system relationships using uniform and mathematical formalism. The SPG approach is useful to build a reactor operator aid system. This paper presents the conception of the application of the SPG approach to the decomposition of TRIGA RC-1 dynamics and for designing of TRIGA diagnostic algorithms. (author)

  11. Job Analysis, Job Descriptions, and Performance Appraisal Systems.

    Science.gov (United States)

    Sims, Johnnie M.; Foxley, Cecelia H.

    1980-01-01

    Job analysis, job descriptions, and performance appraisal can benefit student services administration in many ways. Involving staff members in the development and implementation of these techniques can increase commitment to and understanding of the overall objectives of the office, as well as communication and cooperation among colleagues.…

  12. Tank waste remediation system technical baseline summary description

    International Nuclear Information System (INIS)

    Raymond, R.E.

    1998-01-01

    This document is one of the tools used to develop and control the mission work as depicted in the included figure. This Technical Baseline Summary Description document is the top-level tool for management of the Technical Baseline for waste storage operations

  13. System description for DART (Decision Analysis for Remediation Technologies)

    International Nuclear Information System (INIS)

    Nonte, J.; Bolander, T.; Nickelson, D.; Nielson, R.; Richardson, J.; Sebo, D.

    1997-09-01

    DART is a computer aided system populated with influence models to determine quantitative benefits derived by matching requirements and technologies. The DART database is populated with data from over 900 DOE sites from 10 Field Offices. These sites are either source terms, such as buried waste pits, or soil or groundwater contaminated plumes. The data, traceable to published documents, consists of site-specific data (contaminants, area, volume, depth, size, remedial action dates, site preferred remedial option), problems (e.g., offsite contaminant plume), and Site Technology Coordinating Group (STCG) need statements (also contained in the Ten-Year Plan). DART uses this data to calculate and derive site priorities, risk rankings, and site specific technology requirements. DART is also populated with over 900 industry and DOE SCFA technologies. Technology capabilities can be used to match technologies to waste sites based on the technology''s capability to meet site requirements and constraints. Queries may be used to access, sort, roll-up, and rank site data. Data roll-ups may be graphically displayed

  14. Analysis of Depressurization Performance in Containment of Wolsong NPP Unit 1 through Containment Filtered Venting System

    International Nuclear Information System (INIS)

    Lee, Sunghan; Kim, Jinhyuck; Suh, Nam Duk; Cho, Songwon

    2014-01-01

    Containment filtered venting system (CFVS) is designed to open and to close isolation valves passively by an operator. CFVS is operated when the containment pressure exceeds the design pressure (225 kPa(a)) and is closed when the containment pressure decreases below 151 kPa(a). The aim of this study is to analyze the depressurization performance of Wolsong unit 1 through CFVS during SBO. The thermal-hydraulic behavior in containment of Wolsong unit 1 was evaluated using the MELCOR 1.8.6 code developed at Sandia National Laboratories (SNL) for the U.S. Nuclear Regulatory Commission (NRC). In addition, in order to evaluate the effects of the CFVS according to the venting area, a sensitivity study depending on different venting area of the CFVS was conducted. Finally, an analysis of the effects of filtering and scrubbing of radioactive material for CFVS is important but not treated in this paper. The SBO accident is chosen to analyze the thermal-hydraulic behavior of Wolsong unit 1. During SBO, the analysis of CFVS affecting on the depressurization of the containment was conducted using MELCOR 1.8.6 code. Also, a sensitivity study was carried out to evaluate the depressurization performance according to the venting area of CFVS. The results show that the containment pressure is considerably decreased and the integrity of the containment could be maintained in case of CFVS operating. Therefore, CFVS has the capacity to keep the containment pressure below the design pressure during SBO. In addition, there are large differences in the containment pressure depending on venting area. We found that the decreasing rate of the pressure in the containment and water level in CFVS depends on the venting area. In the future, a proper requirement for CFVS sizing criteria according to accident scenarios such as LBLOCA, SBLOCA and SGTR, etc. should be evaluated in order to review the licensing for CFVS. Finally, analyses of aerosols, fission product, and radioactive material

  15. Permanent monitoring of containment integrity: the sexten system

    International Nuclear Information System (INIS)

    Germain, J.L.; Janneteau, E.

    1993-01-01

    Reactor containment integrity is of prime importance to the safety of PWR units. It is checked by means of tests performed at high pressure during the containment building pressure tests. These periodical tests are supplemented in France by permanent monitoring using the SEXTEN system. First feasibility tests for this system were carried out in 1980. The encouraging results obtained led to the development of a prototype, followed by an industrial system which has since been installed in all French PWR units. This system measures the containment leak rate, with corrections for the compressed air intakes used by the air-operated valves. Leaktightness is expressed in terms of the leak rate for a 60 mbar overpressure. If the leak rate exceeds a fixed limit value, leak detection operations are initiated, using SEXTEN. A new version of the system, known as SEXTEN 2 is being developed. (authors). 2 figs

  16. Code for calculation of spreading of radioactivity in reactor containment systems

    International Nuclear Information System (INIS)

    Vertes, P.

    1992-09-01

    A detailed description of the new version of TIBSO code is given, with applications for accident analysis in a reactor containment system. The TIBSO code can follow the nuclear transition and the spatial migration of radioactive materials. The modelling of such processes is established in a very flexible way enabling the user to investigate a wide range of problems. The TIBSO code system is described in detail, taking into account the new developments since 1983. Most changes improve the capabilities of the code. The new version of TIBSO system is written in FORTRAN-77 and can be operated both under VAX VMS and PC DOS. (author) 5 refs.; 3 figs.; 21 tabs

  17. Description of the control and safety systems of the RA reactor; Opis sistema za upravljanje i sigurnosnu zastitu RA

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, B; Pesic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

    1962-07-01

    This report contains detailed description and scheme of the control and safety system of the RA reactor. It consists of interconnected five systems: for automated regulation; compensation rods; safety rods; power density measurement device; period meter; automated D{sub 2}O level meter in the core. Automated regulation system is divided into two parts: basic system for reactor operation regime at power from 10kW - 10 MW and precise regulation system for operation at set-up power level up to 10 kW which is used occasionally.

  18. Alternatives for high-level waste forms, containers, and container processing systems

    International Nuclear Information System (INIS)

    Crawford, T.W.

    1995-01-01

    This study evaluates alternatives for high-level waste forms, containers, container processing systems, and onsite interim storage. Glass waste forms considered are cullet, marbles, gems, and monolithic glass. Small and large containers configured with several combinations of overpack confinement and shield casks are evaluated for these waste forms. Onsite interim storage concepts including canister storage building, bore holes, and storage pad were configured with various glass forms and canister alternatives. All favorable options include the monolithic glass production process as the waste form. Of the favorable options the unshielded 4- and 7-canister overpack options have the greatest technical assurance associated with their design concepts due to their process packaging and storage methods. These canisters are 0.68 m and 0.54 m in diameter respectively and 4.57 m tall. Life-cycle costs are not a discriminating factor in most cases, varying typically less than 15 percent

  19. Over facility design description for the CPDF [Centrifuge Plant Demonstration Facility]: SDD-1 [System Design Description

    International Nuclear Information System (INIS)

    1987-04-01

    The Centrifuge Plant Demonstration Facility (CPDF) is an essential part of the continuing development of first-production-plant centrifuge technology that will integrate centrifuge machines into a process and enrichment plant design. The CPDF will provide facilities for testing and continued development of a unit cascade in direct support of the commercial Gas Centrifuge Enrichment Plant (GCEP). The basic cascade-oriented equipment, feed, withdrawal, drive system, process piping, utility piping, and other auxiliary and support equipment will be tested in an operating configuration that represents, to the extent possible, GCEP arrangement and operating conditions. The objective will be to demonstrate procedures for production cascade installation, start-up, operation, and maintenance, and to provide proof of overall cascade and associated system design, construction, and operating and maintenance concepts. To the maximum possible extent, all equipment for the CPDF will be procured from commercial sources. Centrifuges will be procured from industry using government-supplied specifications and drawings. The existing Component Preparation Laboratory (CPL) located near the CPDF site will be used for centrifuge component receiving, inspection, assembly, and qualification testing of pre-production test machines. Later in the test program, samples of production machines planned for use in the GCEP will be tested in the CPDF

  20. HUMOS monitoring system of leaks into the containment atmosphere

    International Nuclear Information System (INIS)

    Matal, O.; Zaloudek, J.; Matal, O. Jr.; Klinga, J.; Brom, J.

    1997-01-01

    The detection and monitoring of coolant leaks into the containment atmosphere during reactor operation is a major safety measure. Using the HUMOS monitoring system, leaks can be detected in pressure tests of integrity and in any other mode of operation when the reactor ventilation system is operating and the primary circuit and its components are pressurized. Performance tests, the design, hardware and software of the HUMOS system are briefly described. A test was performed to demonstrate that a small amount of humidity released by leakage into the containment air can be detected. (M.D.)

  1. Information managing in 60Co container inspection system

    International Nuclear Information System (INIS)

    Wu Zhifang; Gu Bohua; Zhou Liye; An Jigang; Liu Yisi

    1998-01-01

    The design, maintenance and realization of information managing database in 60 Co container inspection system made by INET of Tsinghua University is introduced. The technique of Open Database Connectivity (ODBC) is adopted to develop a general format database including text and graphic information. The database application is developed with Visual C ++ 5.0 programming language to run in 32-bit Windows operation system. It conforms to Client/Server model and supports network communication. It works very well in the laboratory emulator of 60 Co container inspection system

  2. The development of the thermohydraulic analysis code for the passive containment cooling system: PCCSAC

    International Nuclear Information System (INIS)

    Wang Jianyu; Zhang Shenru; Min Yuanyou

    1994-01-01

    To estimate the performance of the passive containment cooling system (PCCS) of the AC-600 nuclear power plant, the PCCSAC code is developed currently by the jointed efforts between Tsinghua University and NPIC. Different features on the passive behavior of the system and the main components of the containment are considered in the code which is needed by the further AC-600 R and D Program. With a brief description of the AC-600 passive containment cooling system and components, the main thermohydraulic models and numerical scheme used in the PCCSAC code are introduced and the selected results of the verification and the prediction for the performance of the AC-600 passive containment cooling system under LOCA and a steam line break accident are presented to preliminarily demonstrate the applicability and reliability of the PCCSAC model. The current PCCSAC model is conservative and a further 2-D PCCSAC version is under consideration in addition to provide the database for models from some tests associated with the components and systems unique to AC-600 nuclear power plant to meet the requirement of the more realistic modelization for the AC-600 passive containment cooling system. (author)

  3. 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

  4. Thermal hydraulic analysis of BWR containment venting system

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Sharma, Prashant; Paul, U.K.; Gaikwad, Avinash

    2015-01-01

    Installation of additional containment filtered venting system (CFVS) is necessary to depressurize the containment to maintain its mechanical integrity due to over pressurization during severe accident condition. A typical venting system for BWR is modelled using RELAP5 and analysed to investigate the effect of various thermal hydraulic parameters on the operational parameters of the venting system. The venting system consists of piping from the containment to the scrubber tank and exit line from the scrubber tank. The scrubber tank is partially filled with water to enable the scrubbing action to remove the particulate radionuclides from the incoming containment air. The pipe line from the containment is connected to the venturi inlet and the throat of the venturi is open to the scrubber tank water inventory at designed submergence level. The exit of the venturi is open to scrubber tank water. Filters are used in the upper air space of the scrubber tank as mist separator before venting out the air into the atmosphere through the exit vent line. The effect of thermal hydraulic parameters such as inlet fluid temperature, inlet steam content and venturi submergence in the scrubber tank on the venting flow rate, exit steam content, scrubber tank inventory, overflow line and siphon breaker flow rate is analysed. Results show that inlet steam content and the venturi nozzle submergence influence the venting system parameters. (author)

  5. Status of advanced containment systems for next generation water reactors

    International Nuclear Information System (INIS)

    1994-06-01

    The present IAEA status report is intended to provide information on the current status and development of containment systems of the next generation reactors for electricity production and, particularly, to highlight features which may be considered advanced, i.e. which present improved performance with evolutionary or innovative design solutions or new design approaches. The objectives of the present status report are: To present, on a concise and consistent basis, selected containment designs currently being developed in the world; to review and compare new approaches to the design bases for the containments, in order to identify common trends, that may eventually lead to greater worldwide consensus, to identify, list and compare existing design objectives for advanced containments, related to safety, availability, maintainability, plant life, decommissioning, economics, etc.; to describe the general approaches adopted in different advanced containments to cope with various identified challenges, both those included in the current design bases and those related to new events considered in the design; to briefly identify recent achievements and future needs for new or improved computer codes, standards, experimental research, prototype testing, etc. related to containment systems; to describe the outstanding features of some containments or specific solutions proposed by different parties and which are generally interesting to the international scientific community. 36 refs, 27 figs, 1 tab

  6. System of two containers contaminated on the inside

    International Nuclear Information System (INIS)

    Hager, L.; Heller, G.

    1983-01-01

    Two lids coupled together of a system of two containers contaminated on the inside form a frustrum of a cone with an outer surface decreasing smoothly in the same direction. The seats for these lids form two openings of the containers of the frustrum of a cone-shaped jacket matched to the jacket of the two coupled lids. The outsides of the two lids and the outsides of the containers form a surface in the same plane as the openings in the annular regions of at least one of the jacket surfaces of the frustrum of a cone or the seats of the frustrum of a cone jacket. (orig./HP) [de

  7. ACE puts containment venting systems to the test

    International Nuclear Information System (INIS)

    Merilo, M.

    1990-01-01

    Filtered venting of reactor containments has received considerable attention recently as a method for avoiding containment failure due to overpressure during severe accidents. Several proposed filtration devices have been tested in the internationally sponsored Advanced Containment Experiments (ACE) programme, such that a self consistent comparison of the aerosol removal characteristics of these systems could be obtained. Considering the different design, requirements and operating conditions of the filter devices, a direct comparison is not possible, nor appropriate. Nevertheless, large scale models, using full scale elements of the various devices whenever feasible, have been tested with consistent mixtures of aerosols and carrier gases. (author)

  8. Emergency air cleaning system development for LMFBR containments

    International Nuclear Information System (INIS)

    McCormack, J.D.; Hilliard, R.K.; Postma, A.K.; Muhlestein, L.D.

    1975-01-01

    Criteria for evaluating the various types of Emergency Air Cleaning Systems which may be used in LMFBR plants have been established for both single containment and containment-confinement arrangements. These two plant arrangements have quite different air cleaning requirements for postulated design base accident conditions. Work is currently in progress to select from a list of candidate air cleaning systems those which best meet the criteria requirements. By means of a weighted rating system, areas of strength or weakness can be found and the conceptual system design then optimized. The final system arrangements will be ranked and several of the most promising systems selected for large-scale tests in the former CSE vessel at Hanford. 8 references. (U.S.)

  9. Study on vent containment filtering for the Spanish NPPS systems

    International Nuclear Information System (INIS)

    Peinado, A.; Serrano, C.; Garcia-Serrano, J. L.

    2013-01-01

    The study discusses filtering systems on the market, and its suppliers, taking into account aspects such as ease of integration into the current plant design, characteristics of the process of filtering, operational range, autonomy of the system, maintenance, qualification and proven experiences, among others. The study, also contains an analysis of sequences kind of accident that serve to define the design parameters of the system.

  10. Development of Wireless System for Containment Integrated Leakage Rate Test

    International Nuclear Information System (INIS)

    Lee, Kwang-Dae; Oh, Eung-Se; Yang, Seung-Ok

    2006-01-01

    The containment system leakage rate should be estimated periodically with reliable test equipment. In light-water reactor nuclear power plants, ANSI/ANS- 56.8 is a basis for determining leakage rates. Two types of data acquisition system, centralized type and networked type, has been used. In centralized type, all sensors are connected directly from sensors in the containment to the measuring equipment outside the building. The other hand, the networked type has several branch chains which connect one group of the network-sensors together. To test leakage rate, more than 20 temperature sensors and 6 humidity sensors, which are different for each plant, should be installed on a specific level in the containment. A wireless technology gives the benefits such as reducing installation efforts, making pretest easy, so it is widely used more and more in the plant monitoring. As the containment system has many kinds of complex barriers to the radio frequency, the radio power and frequency band for better transmission rate as well as the interference by the radio frequency should be considered. The overview of the wireless sensor system for the containment leakage rate test is described here and the test results on Yonggwang unit 4 PWR plant is presented

  11. Engineer/constructor description of work for Tank 241-SY-102 retrieval system, project W-211, initial tank retrieval systems

    International Nuclear Information System (INIS)

    Rieck, C.A.

    1996-02-01

    This document provides a description of work for the design and construction of a waste retrieval system for Tank 241-SY-102. The description of work includes a working estimate and schedule, as well as a narrative description and sketches of the waste retrieval system. The working estimate and schedule are within the established baselines for the Tank 241-SY-102 retrieval system. The technical baseline is provided in Functional Design Criteria, WHC-SD-W211-FDC-001, Revision 2

  12. CLASSIFICATION OF THE MGR DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER SYSTEM

    International Nuclear Information System (INIS)

    J.A. Ziegler

    1999-01-01

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

  13. System for cooling the containment vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

    The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

  14. Corrosion of copper alloys in sulphide containing district heting systems

    DEFF Research Database (Denmark)

    Thorarinsdottir, R.I.; Maahn, Ernst Emanuel

    1999-01-01

    Copper and some copper alloys are prone to corrosion in sulphide containing geothermal water analogous to corrosion observed in district heating systems containing sulphide due to sulphate reducing bacteria. In order to study the corrosion of copper alloys under practical conditions a test...... was carried out at four sites in the Reykjavik District Heating System. The geothermal water chemistry is different at each site. The corrosion rate and the amount and chemical composition of deposits on weight loss coupons of six different copper alloys are described after exposure of 12 and 18 months......, respectively. Some major differences in scaling composition and the degree of corrosion attack are observed between alloys and water types....

  15. A thermodynamic description of the system Pd-Rh-H-D-T

    Energy Technology Data Exchange (ETDEWEB)

    Joubert, J.-M., E-mail: jean-marc.joubert@icmpe.cnrs.fr [Chimie Metallurgique des Terres Rares, Institut de Chimie et des Materiaux Paris-Est, CNRS, Universite Paris-Est, UMR 7182, 2-8 Rue Henri Dunant, F-94320 Thiais (France); Thiebaut, S. [CEA/DAM/Valduc, F-21120 Is sur Tille (France)

    2011-02-15

    The quinary system D-H-Pd-Rh-T has been described thermodynamically by the CALPHAD approach. Previous descriptions of the binary subsystems have been used. To model the high pressure data an equation of state for the gases D{sub 2} and T{sub 2} compatible with the CALPHAD approach has been obtained similar to that previously used for H{sub 2}. A complete literature search has been undertaken for the three ternary systems H-Pd-Rh, D-Pd-Rh and Pd-Rh-T and the most significant experimental data have been selected for a thermodynamic assessment of these systems. In order to complement the available data, pressure-composition curves have been measured at different temperatures for the two last systems in the present work. Calculations and optimization of the system under para-equilibrium conditions, i.e. in pseudo-binary systems (Pd,Rh)-H, (Pd,Rh)-D or (Pd,Rh)-T, have been achieved using a pseudo-atom describing the Pd-Rh solid solution. This special method allows the presence of a miscibility gap in the binary metallic system to be dealt with. We show that a simple combination of the binary systems alone is unable to properly describe these ternary systems and that ternary interaction parameters have to be introduced. The binary and ternary systems may then be combined to perform calculations in the quinary D-H-Pd-Rh-T system. It is believed that extrapolation in systems containing different isotopes are fairly accurate provided that the so-called Toop model is used.

  16. Histological description of Cerdocyon thous (Linnaeus, 1766 respiratory system

    Directory of Open Access Journals (Sweden)

    Marla P. Rocha

    Full Text Available ABSTRACT: The massive agricultural expansion converted the Cerdocyon thous, a South American native predator, in vulnerable specie. Basic data, such as histological description, are important to raise awareness on animal species, helping on preservation strategies. Considering the difficult in obtain samples, as the euthanasia of wild animals for this purpose is not allowed, data on histology are very scarce or inexistent. The objective of this paper was to provide a detailed histological description of the trachea and bronchial tree of the crab-eating fox Cerdocyon thous (Linnaeus, 1766. The specimens (one adult male and one adult female used were provided by the Federal University of Pelotas (Pelotas, RS, Brazil Rehabilitation Center of Wild Fauna (NURFS. Tissue samples were fixed in 10% formalin and included in paraffin. After slicing, samples were stained with HE (hematoxylin and eosin, PAS (periodic acid-Schiff and resorcin fuchsin. Trachea had an average diameter of 7.87mm, and approximately 57% of the mucosa ciliated pseudo-stratified columnar epithelium was composed of goblet cells, mostly in the dorsal region. Bronchia and bronchioles had a mucosal fold with higher number of goblet cells. Using all these techniques there is no great remarkable differences from C. thous trachea and lung, when compared with the previous described structures for carnivores and most mammals, except for the goblet cells “regionalization”. Described results are important to understand the animal physiological and behavioral habits, allowing the development of preservation and protection strategies.

  17. APPLICATION OF A GEOGRAPHIC INFORMATION SYSTEM FOR A CONTAINMENT SYSTEM LEAK DETECTION

    Science.gov (United States)

    The use of physical and hydraulic containment systems for the isolation of contaminated ground water associated with hazardous waste sites has increased during the last decade. Existing methodologies for monitoring and evaluating leakage from hazardous waste containment systems ...

  18. Effects of aging in containment spray injection system of PWR reactor containment

    International Nuclear Information System (INIS)

    Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L.

    2014-01-01

    This paper presents a contribution to the study of the components aging process in commercial plants of Pressurized Water Reactors (PWR). The analysis is done by applying the method of Fault trees, Monte Carlo Method and Fussell-Vesely Importance Measurement. The study on the aging of nuclear plants, is related to economic factors involved directly with the extent of their operational life, and also provides important data on issues of safety. The most recent case involving the process of extending the life of a PWR plant can be seen in Angra 1 Nuclear Power Plant by investing $ 27 million in the installation of a new reactor cover. The corrective action generated an extension of the useful life of Angra 1 estimated in twenty years, and a great savings compared to the cost of building a new plant and the decommissioning of the first, if it had reached the operation time out 40 years. The extension of the lifetime of a nuclear power plant must be accompanied by special attention from the most sensitive components of the systems to the aging process. After the application of the methodology (aging analysis of Containment Spray Injection System (CSIS)) proposed in this paper, it can be seen that increasing the probability of failure of each component, due to the aging process, generate an increased general unavailability of the system that contains these basic components. The final results obtained were as expected and can contribute to the maintenance policy, preventing premature aging in nuclear power systems

  19. Spent nuclear fuel project cold vacuum drying facility tempered water and tempered water cooling system design description

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1998-01-01

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Tempered Water (TW) and Tempered Water Cooling (TWC) System . The SDD was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), The HNF-SD-SNF-DRD-O02, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the TW and TWC equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SOD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  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. System Design Description and Requirements for Modeling the Off-Gas Systems for Fuel Recycling Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Daryl R. Haefner; Jack D. Law; Troy J. Tranter

    2010-08-01

    This document provides descriptions of the off-gases evolved during spent nuclear fuel processing and the systems used to capture the gases of concern. Two reprocessing techniques are discussed, namely aqueous separations and electrochemical (pyrochemical) processing. The unit operations associated with each process are described in enough detail so that computer models to mimic their behavior can be developed. The document also lists the general requirements for the desired computer models.

  3. Numerical Analysis of a Passive Containment Filtered Venting System

    International Nuclear Information System (INIS)

    Kim, Taejoon; Ha, Huiun; Heo, Sun

    2014-01-01

    The passive Containment Filtered Venting system (CFVS) does not have principally any kind of isolation valves or filtering devices which need periodic maintenance. In this study, the hydro-thermal analysis is presented to investigate the existence of flow instability in the passive CFVS and its performance under the pressure change of APR+ containment building with LB-LOCA M/E data. The Passive Containment Filtered Venting System was suggested as a part in i-Power development project and the operation mechanism was investigated by numerical modeling and simulation using GOTHIC8.0 system code. There are four Phases for consideration to investigate the pressurization of the containment building, loss of hydrostatic head in the pipe line of CFVS, opening of pipe line and gas ejection to the coolant tank, and the head recovery inside the pipe as the containment gas exhausted. The simulation results show that gas generation rate determine the timing of head recovery in the CFVS pipe line and that the equipment of various devices inducing pressure loss at the pipe can give the capacity of Phase control of the passive CFVS operation

  4. Cold Vacuum Drying facility potable water system design description

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) potable water (PW) system. The PW system provides potable water to the CVDF for supply to sinks, water closets, urinals, showers, custodial service sinks, drinking fountains, the decontamination shower, supply water to the non-PW systems, and makeup water for the de-ionized water system

  5. Hardware descriptions of the I and C systems for NPP

    International Nuclear Information System (INIS)

    Lee, Cheol Kwon; Oh, In Suk; Park, Joo Hyun; Kim, Dong Hoon; Han, Jae Bok; Shin, Jae Whal; Kim, Young Bak

    2003-09-01

    The hardware specifications for I and C Systems of SNPP(Standard Nuclear Power Plant) are reviewed in order to acquire the hardware requirement and specification of KNICS (Korea Nuclear Instrumentation and Control System). In the study, we investigated hardware requirements, hardware configuration, hardware specifications, man-machine hardware requirements, interface requirements with the other system, and data communication requirements that are applicable to SNP. We reviewed those things of control systems, protection systems, monitoring systems, information systems, and process instrumentation systems. Through the study, we described the requirements and specifications of digital systems focusing on a microprocessor and a communication interface, and repeated it for analog systems focusing on the manufacturing companies. It is expected that the experience acquired from this research will provide vital input for the development of the KNICS

  6. B-Plant Canyon Ventilation Control System Description; FINAL

    International Nuclear Information System (INIS)

    MCDANIEL, K.S.

    1999-01-01

    Project W-059 installed a new B Plant Canyon Ventilation System. Monitoring and control of the system is implemented by the Canyon Ventilation Control System (CVCS). This document describes the CVCS system components which include a Programmable Logic Controller (PLC) coupled with an Operator Interface Unit (OIU) and application software. This document also includes an Alarm Index specifying the setpoints and technical basis for system analog and digital alarms

  7. Process and closure system for a radioactive waste container

    International Nuclear Information System (INIS)

    Meyer, Andre.

    1974-01-01

    The closure process described is for a cylindrical radioactive waste drum. It makes use of a closure system for the drum comprising two lids separated by a twin flange seal. It consists essentially in placing a double flange 'O' ring inside the upper lip of the drum, and after filling has been completed, fitting the first lid on the twin flange 'O' ring and pushing down this lid whilst squashing the upper flange and then putting on the second lid in the usual prescribed manner. A description is also given of the drum sealing apparatus [fr

  8. The mathematical description of resonances in many-body systems

    International Nuclear Information System (INIS)

    Orth, A.

    1985-01-01

    We introduce a characterization for quantum-mechanical resonance and use it in order to detect for certain distinct physical states an especially slow decay behaviour. We apply these results to a model of the quantum-mechanical many-body problem and obtain so a mathematical description of the Auger effect (self-ionization of atoms). The class of the interaction potentials admitted for our theory is compared with other theories on resonances extremely large. We establish differentiability conditions and conditions on the fading behaviour in the infinite. Especially the Coulomb potential and the Yukawa potential belong to our class but also non-spherical-symmetric and non-analytic potentials with a Coulomb-like singularity in the origin, two- to threefold differentiable which tend to zero at the infinite. In the introduction we discuss extensively also by means of some examples the problematics of the quantum-mechanical resonance. (orig.) [de

  9. International safeguards data management system. System description: Version 1.1.; Release PLI 82

    International Nuclear Information System (INIS)

    Argentesi, T.; Casilli, T.; Costantini, L.; Dondi, M.G.; Franklin, M.

    1982-01-01

    This document describes a nuclear material accountancy system which has been developed using the ADABAS data base management system and is implemented on the JRC-Ispra computer. Throughout the report, the data base system is referred to as ''ISADAM'', i.e. International Safeguards Data Management System. The system provides tools for a safeguards authority to decide whether an operators MUF (Material Unaccounted For) can be accounted for as an accumulation of operator measurement errors. The principle objective of the applications programs described here is to provide a variance analysis of MUF in which the variance of MUF is computed as a function of the accounting declarations and the error characteristics of the operator measurement system. A overview of ISADAM is presented; then, a detailed description of the processing applied by the system is given. A description of the parameter information required by the four autonomous programs ISADAM is presented. In developing ISADAM, one of the prime factors taken into consideration was the ease with which it could be used

  10. Development of an accident management expert system for containment assessment

    International Nuclear Information System (INIS)

    Nelson, W.R.; Sebo, D.E.; Haney, L.N.

    1987-01-01

    The United States Nuclear Regulatory Commission (USNRSC) is sponsoring a program at the Idaho National Engineering Laboratory (INEL) to develop an accident management expert system. The intended users of the system are the personnel of the NRC Operations Center in Washington, D.C. The expert system will be used to help NRC personnel monitor and evaluate the status and management of the containment during a severe reactor accident. The knowledge base will include severe accident knowledge regarding the maintenance of the critical safety functions, especially containment integrity, during an accident. This paper summarizes the concepts that have been developed for the accident management expert system, and the plans that have been developed for its implementation

  11. Qualitative Description of Electric Power System Future States

    Energy Technology Data Exchange (ETDEWEB)

    Hardy, Trevor D.; Corbin, Charles D.

    2018-03-06

    The simulation and evaluation of transactive systems depends to a large extent on the context in which those efforts are performed. Assumptions regarding the composition of the electric power system, the regulatory and policy environment, the distribution of renewable and other distributed energy resources (DERs), technological advances, and consumer engagement all contribute to, and affect, the evaluation of any given transactive system, regardless of its design. It is our position that the assumptions made about the state of the future power grid will determine, to some extent, the systems ultimately deployed, and that the transactive system itself may play an important role in the evolution of the power system.

  12. Improving the performance of sorter systems by scheduling inbound containers

    NARCIS (Netherlands)

    Haneyah, S.W.A.; Schutten, Johannes M.J.; Fikse, K.

    2013-01-01

    This paper addresses the inbound containers scheduling problem for automated sorter systems in two different industrial sectors: parcel & postal sorting and baggage handling. We build on existing literature, particularly on the dynamic load balancing algorithm designed for the parcel hub scheduling

  13. Structure-rheology relations in sodium caseinate containing systems

    NARCIS (Netherlands)

    Ruis, H.G.M.

    2007-01-01

    The general aim of the work described in this thesis was to investigate structure-rheologyrelations for dairy related products, focusing on model systems containing sodium caseinate. The acid inducedgelationof sodium caseinate, of sodium caseinate stabilized emulsions, and the effect of shear on the

  14. Safety of systems for the retention of wastes containing radionuclides

    International Nuclear Information System (INIS)

    1980-11-01

    Information and minimal requirements demanded by CNEN for the emission of the Approval Certificate of the Safety Analysis Report related to system for the retention of wastes containing radionuclide, are established, aiming to assure low radioactivity levels to the environment. (E.G.) [pt

  15. Cold Vacuum Drying facility heating, ventilation, and Air Conditioning system design description

    International Nuclear Information System (INIS)

    SINGH, G.

    2000-01-01

    This System Design Description (SDD) addresses the HVAC system for the CVDF. The CVDF HVAC system consists of five subsystems: (1) Administration building HVAC system; (2) Process bay recirculation HVAC system; (3) Process bay local exhaust HVAC and process vent system; (4) Process general supply/exhaust HVAC system; and (5) Reference air system. The HVAC and reference air systems interface with the following systems: the fire protection control system, Monitoring and Control System (MCS), electrical power distribution system (including standby power), compressed air system, Chilled Water (CHW) system, drainage system, and other Cold Vacuum Drying (CVD) control systems not addressed in this SDD

  16. Cold Vacuum Drying facility personnel monitoring system design description

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) instrument air (IA) system that provides instrument quality air to the CVDF. The IA system provides the instrument quality air used in the process, HVAC, and HVAC instruments. The IA system provides the process skids with air to aid in the purging of the annulus of the transport cask. The IA system provides air for the solenoid-operated valves and damper position controls for isolation, volume, and backdraft in the HVAC system. The IA system provides air for monitoring and control of the HVAC system, process instruments, gas-operated valves, and solenoid-operated instruments. The IA system also delivers air for operating hand tools in each of the process bays

  17. Cold Vacuum Drying facility fire protection system design description

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) fire protection system (FPS). The FPS provides fire detection, suppression, and loss limitation for the CVDF structure, personnel, and in-process spent nuclear fuel. The system provides, along with supporting interfacing systems, detection, alarm, and activation instrumentation and controls, distributive piping system, isolation valves, and materials and controls to limit combustibles and the associated fire loadings

  18. Cold Vacuum Drying facility sanitary sewage collection system design description

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) sanitary sewage collection system. The sanitary sewage collection system provides collection and storage of effluents and raw sewage from the CVDF to support the cold vacuum drying process. This system is comprised of a sanitary sewage holding tank and pipes for collection and transport of effluents to the sanitary sewage holding tank

  19. Description and Documentation of the Dental School Dental Delivery System.

    Science.gov (United States)

    Chase, Rosen and Wallace, Inc., Alexandria, VA.

    A study was undertaken to describe and document the dental school dental delivery system using an integrated systems approach. In late 1976 and early 1977, a team of systems analysts and dental consultants visited three dental schools to observe the delivery of dental services and patient flow and to interview administrative staff and faculty.…

  20. Personnel protection and beam containment systems for the 3 GeV Injector

    International Nuclear Information System (INIS)

    Yotam, R.; Cerino, J.; Garoutte, R.; Hettel, R.; Horton, M.; Sebek, J.; Benson, E.; Crook, K.; Fitch, J.; Ipe, N.; Nelson, G.; Smith, H.

    1991-01-01

    The 3 GeV Injector is the electron beam source for the SPEAR Storage Ring, and its personnel safety system was designed to protect personnel from both radiation exposure and electrical hazards. The Personnel Protection System (PPS) was designed and implemented with complete redundancy and is a relay based interlock system completely independent from the machine protection system. A comprehensive monitoring of the system status, and control of the Injector PPS from the SPEAR Control Room via the control computer is a feature. The Beam Containment System (BCS) is based on beam current measurements along the Linac and on Beam Shut Off Ion Chambers (BSOIC) installed outside the Linac, at several locations around the Booster, and around the SPEAR storage ring. An outline of the design criteria is presented with more detailed description of the philosophy of the PPS logic and the BCS

  1. System Description for Tank 241-AZ-101 Waste Retrieval Data Acquisition System

    International Nuclear Information System (INIS)

    ROMERO, S.G.

    2000-01-01

    The proposed activity provides the description of the Data Acquisition System for Tank 241-AZ-101. This description is documented in HNF-5572, Tank 241-AZ-101 Waste Retrieval Data Acquisition System (DAS). This activity supports the planned mixer pump tests for Tank 241-AZ-101. Tank 241-AZ-101 has been selected for the first full-scale demonstration of a mixer pump system. The tank currently holds over 960,000 gallons of neutralized current acid waste, including approximately 12.7 inches of settling solids (sludge) at the bottom of the tank. As described in Addendum 4 of the FSAR (LMHC 2000a), two 300 HP mixer pumps with associated measurement and monitoring equipment have been installed in Tank 241-AZ-101. The purpose of the Tank 241-AZ-101 retrieval system Data Acquisition System (DAS) is to provide monitoring and data acquisition of key parameters in order to confirm the effectiveness of the mixer pumps utilized for suspending solids in the tank. The suspension of solids in Tank 241-AZ-101 is necessary for pretreatment of the neutralized current acid waste and eventual disposal as glass via the Hanford Waste Vitrification Plant. HNF-5572 provides a basic description of the Tank 241-AZ-101 retrieval system DAS, including the field instrumentation and application software. The DAS is provided to fulfill requirements for data collection and monitoring. This document is not an operations procedure or is it intended to describe the mixing operation. This USQ screening provides evaluation of HNF-5572 (Revision 1) including the changes as documented on ECN 654001. The changes include (1) add information on historical trending and data backup, (2) modify DAS I/O list in Appendix E to reflect actual conditions in the field, and (3) delete IP address in Appendix F per Lockheed Martin Services, Inc. request

  2. Optimal design of passive containment cooling system for innovative PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon [Central Research Institute, Korea Hydro and Nuclear Power, Ltd., Daejeon (Korea, Republic of)

    2017-08-15

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  3. Optimal design of passive containment cooling system for innovative PWR

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  4. Optimal design of passive containment cooling system for innovative PWR

    International Nuclear Information System (INIS)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon

    2017-01-01

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed

  5. A review on leakage rate tests for containment isolation systems

    International Nuclear Information System (INIS)

    Kim, In Goo; Kim, Hho Jung

    1992-01-01

    Wide experiences in operating containment isolation systems have been accumulated in Korea since 1978. Hence, it becomes necessary to review the operating data in order to confirm the integrity of containments with about 50 reactor-years of experience and to establish the future direction to the containment test program. The objectives of present work are to collect, consolidate and assess the leakage rate data, and then to find out dominant leakage paths and factors affecting integrated leakage rate test. General trends of overall leakage show that more careful surveillance during pre-operational test can reduce the containment leakage. Dominant leakage paths are found to be through air locks and large-sized valves, such as butterfly valves of purge lines, so that weighted surveillance and inspection on these dominant leakage paths can considerably reduce the containment leakage. The atmosphere stabilization are found to be the most important to obtain the reliable result. In order to get well stabilized atmosphere, temperature and flow rate of compressed air should be kept constant and it is preferable not to operate fan cooler during pressurizing the containment for test

  6. Test plan for buried waste containment system materials

    International Nuclear Information System (INIS)

    Weidner, J.; Shaw, P.

    1997-03-01

    The objectives of the FY 1997 barrier material work at the Idaho National Engineering and Environmental Laboratory are to (1) select a waste barrier material and verify that it is compatible with the Buried Waste Containment System Process, and (2) determine if, and how, the Buried Waste Containment System emplacement process affects the material properties and performance (on proof of principle scale). This test plan describes a set of measurements and procedures used to validate a waste barrier material for the Buried Waste Containment System. A latex modified proprietary cement manufactured by CTS Cement Manufacturing Company will be tested. Emplacement properties required for the Buried Waste Containment System process are: slump between 8 and 10 in., set time between 15 and 30 minutes, compressive strength at set of 20 psi minimum, and set temperature less than 100 degrees C. Durability properties include resistance to degradation from carbonate, sulfate, and waste-site soil leachates. A set of baseline barrier material properties will be determined to provide a data base for comparison with the barrier materials when tested in the field. The measurements include permeability, petrographic analysis to determine separation and/or segregation of mix components, and a set of mechanical properties. The measurements will be repeated on specimens from the field test material. The data will be used to determine if the Buried Waste Containment System equipment changes the material. The emplacement properties will be determined using standard laboratory procedures and instruments. Durability of the barrier material will be evaluated by determining the effect of carbonate, sulfate, and waste-site soil leachates on the compressive strength of the barrier material. The baseline properties will be determined using standard ASTM procedures. 9 refs., 1 fig., 2 tabs

  7. 41 CFR 101-29.102 - Use of metric system of measurement in Federal product descriptions.

    Science.gov (United States)

    2010-07-01

    ... PROCUREMENT 29-FEDERAL PRODUCT DESCRIPTIONS 29.1-General § 101-29.102 Use of metric system of measurement in... measurement in Federal product descriptions. 101-29.102 Section 101-29.102 Public Contracts and Property... Federal agencies to: (a) Maintain close liaison with other Federal agencies, State and local governments...

  8. Reliability assessment of passive containment isolation system using APSRA methodology

    International Nuclear Information System (INIS)

    Nayak, A.K.; Jain, Vikas; Gartia, M.R.; Srivastava, A.; Prasad, Hari; Anthony, A.; Gaikwad, A.J.; Bhatia, S.; Sinha, R.K.

    2008-01-01

    In this paper, a methodology known as APSRA (Assessment of Passive System ReliAbility) has been employed for evaluation of the reliability of passive systems. The methodology has been applied to the passive containment isolation system (PCIS) of the Indian advanced heavy water reactor (AHWR). In the APSRA methodology, the passive system reliability evaluation is based on the failure probability of the system to carryout the desired function. The methodology first determines the operational characteristics of the system and the failure conditions by assigning a predetermined failure criterion. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the PCIS performance. APSRA proposes to compare the code predictions with the test data to generate the uncertainties on the failure parameter prediction, which is later considered in the code for accurate prediction of failure surface of the system. Once the failure surface of the system is predicted, the cause of failure is examined through root diagnosis, which occurs mainly due to failure of mechanical components. The failure probability of these components is evaluated through a classical PSA treatment using the generic data. The reliability of the PCIS is evaluated from the probability of availability of the components for the success of the passive containment isolation system

  9. Quantum description of microscopic and macroscopic systems: Old problems and recent investigations

    International Nuclear Information System (INIS)

    Ghirardi, G.C.

    1986-04-01

    We review some open problems and some proposed solutions which are encountered in the quantum description of the microscopic systems, of the macroscopic ones, and of the interactions between these two types of objects. We describe a recent attempt allowing a unified description of all phenomena, reproducing the quantum mechanical situation for microscopic systems and inducing in a completely consistent way the classical behaviour of macro object and the phenomena of wave packet reduction in the system-apparatus interaction. (author)

  10. Multiple Description Coding for Closed Loop Systems over Erasure Channels

    DEFF Research Database (Denmark)

    Østergaard, Jan; Quevedo, Daniel

    2013-01-01

    In this paper, we consider robust source coding in closed-loop systems. In particular, we consider a (possibly) unstable LTI system, which is to be stabilized via a network. The network has random delays and erasures on the data-rate limited (digital) forward channel between the encoder (controller......) and the decoder (plant). The feedback channel from the decoder to the encoder is assumed noiseless. Since the forward channel is digital, we need to employ quantization.We combine two techniques to enhance the reliability of the system. First, in order to guarantee that the system remains stable during packet...... by showing that the system can be cast as a Markov jump linear system....

  11. Analytic descriptions of stochastic bistable systems under force ramp.

    Science.gov (United States)

    Friddle, Raymond W

    2016-05-01

    Solving the two-state master equation with time-dependent rates, the ubiquitous driven bistable system, is a long-standing problem that does not permit a complete solution for all driving rates. Here we show an accurate approximation to this problem by considering the system in the control parameter regime. The results are immediately applicable to a diverse range of bistable systems including single-molecule mechanics.

  12. Cold Vacuum Drying facility deionized water system design description

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) de-ionized water system. The de-ionized water system is used to provide clean, conditioned water, free from contaminants, chlorides and iron for the CVD Facility. Potable water is supplied to the deionized water system, isolated by a backflow prevention device. After the de-ionization process is complete, via a packaged de-ionization unit, de-ionized water is supplied to the process deionization unit

  13. POOL WATER TREATMENT AND COOLING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    King, V.

    2000-01-01

    The Pool Water Treatment and Cooling System is located in the Waste Handling Building (WHB), and is comprised of various process subsystems designed to support waste handling operations. This system maintains the pool water temperature within an acceptable range, maintains water quality standards that support remote underwater operations and prevent corrosion, detects leakage from the pool liner, provides the capability to remove debris from the pool, controls the pool water level, and helps limit radiological exposure to personnel. The pool structure and liner, pool lighting, and the fuel staging racks in the pool are not within the scope of the Pool Water Treatment and Cooling System. Pool water temperature control is accomplished by circulating the pool water through heat exchangers. Adequate circulation and mixing of the pool water is provided to prevent localized thermal hotspots in the pool. Treatment of the pool water is accomplished by a water treatment system that circulates the pool water through filters, and ion exchange units. These water treatment units remove radioactive and non-radioactive particulate and dissolved solids from the water, thereby providing the water clarity needed to conduct waste handling operations. The system also controls pool water chemistry to prevent advanced corrosion of the pool liner, pool components, and fuel assemblies. Removal of radioactivity from the pool water contributes to the project ALARA (as low as is reasonably achievable) goals. A leak detection system is provided to detect and alarm leaks through the pool liner. The pool level control system monitors the water level to ensure that the minimum water level required for adequate radiological shielding is maintained. Through interface with a demineralized water system, adequate makeup is provided to compensate for loss of water inventory through evaporation and waste handling operations. Interface with the Site Radiological Monitoring System provides continuous

  14. System design description for the LDUA common video end effector system (CVEE)

    International Nuclear Information System (INIS)

    Pardini, A.F.

    1998-01-01

    The Common Video End Effector System (CVEE), system 62-60, was designed by the Idaho National Engineering Laboratory (INEL) to provide the control interface of the various video end effectors used on the LDUA. The CVEE system consists of a Support Chassis which contains the input and output Opto-22 modules, relays, and power supplies and the Power Chassis which contains the bipolar supply and other power supplies. The combination of the Support Chassis and the Power Chassis make up the CVEE system. The CVEE system is rack mounted in the At Tank Instrument Enclosure (ATIE). Once connected it is controlled using the LDUA supervisory data acquisition system (SDAS). Video and control status will be displayed on monitors within the LDUA control center

  15. NUDAT. System for access to nuclear data. Summary description

    International Nuclear Information System (INIS)

    Dunford, C.L.; Kinsey, R.R.

    1998-01-01

    The NUDAT program with its associated database provides access to nuclear properties and some nuclear reaction data. The program has interfaces for WWW, Telnet online access, and PC. The database contains the following information: level and gamma-ray adopted properties from ENSDF; nuclear ground and metastable state properties; radioactive decay radiations from ENSDF; thermal neutron cross sections and resonance integrals as published in 'Neutron Cross Sections', Vol. 1. The online version is accessible through the IAEA's WWW site or through the Telnet online service NDIS, the PC version is available by FTP or on CD-ROM. (author)

  16. Geometric Description of Fibre Bundle Surface for Birkhoff System

    International Nuclear Information System (INIS)

    Li-Mei, Cao; Hua-Fei, Sun; Zhen-Ning, Zhang

    2009-01-01

    A fibre bundle surface for the Birkhoff system is constructed. The metric and the Riemannian connection of the surface are defined and the representation of the Gaussian curvature of this surface is presented. Finally, three examples for the Birkhoff system are given to illustrate our results. (general)

  17. Cold Vacuum Drying (CVD) Electrical System Design Description

    International Nuclear Information System (INIS)

    BRISBIN, S.A.

    1999-01-01

    This document provides a technical explanation of the design and operation of the electrical system for the Cold Vacuum Drying Facility. This document identifies the requirements, and the basis for the requirements and details on how the requirements have been implemented in the design and construction of the facility. This document also provides general guidance for the surveillance, testing, and maintenance of this system

  18. Summary description of the scale modular code system

    International Nuclear Information System (INIS)

    Parks, C.V.

    1987-12-01

    SCALE - a modular code system for Standardized Computer Analyses for Licensing Evaluation - has been developed at Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission staff. The SCALE system utilizes well-established computer codes and methods within standard analytic sequences that allow simplified free-form input, automate the data processing and coupling between codes, and provide accurate and reliable results. System development has been directed at criticality safety, shielding, and heat transfer analysis of spent fuel transport and/or storage casks. However, only a few of the sequences (and none of the individual functional modules) are restricted to cask applications. This report will provide a background on the history of the SCALE development and review the components and their function within the system. The available data libraries are also discussed, together with the automated features that standardize the data processing and systems analysis. 83 refs., 32 figs., 11 tabs

  19. High temperature performance limit of containment system of transport cask

    International Nuclear Information System (INIS)

    Kato, Osamu; Saegusa, Toshiari

    1998-01-01

    The containment performance of a containment system using elastomer gaskets for transport casks under a high temperature and high pressure was clarified. Major results are as follows; (1) The deformation characteristics of the gaskets were represented by the compressive permanent strain rate (Dp). The temperature and time dependence was shown by Larson-Miller Parameter (LMP). (2) Generally, the high temperature performance limit is obtained by a value of LMP when the Dp value reaches 80%. However, the gaskets (FKM, VMQ, EPDM) used for real transport casks were not damaged and the containment performance was not deteriorated as a conservative condition. (3) Assuming that the service period of the gaskets for transport casks is 3 months or 1 year, the high temperature performance limit of the gasket made of fluorine rubber (FKM) is 202degC or 182degC, respectively, which includes safety margin. (author)

  20. A PC-based discrete tomography imaging software system for assaying radioactive waste containers

    International Nuclear Information System (INIS)

    Palacios, J.C.; Longoria, L.C.; Santos, J.; Perry, R.T.

    2003-01-01

    A PC-based discrete tomography imaging software system for assaying radioactive waste containers for use in facilities in Mexico has been developed. The software system consists of three modules: (i) for reconstruction transmission tomography, (ii) for reconstruction emission tomography, and (iii) for simulation tomography. The Simulation Module is an interactive computer program that is used to create simulated databases for input to the Reconstruction Modules. These databases may be used in the absence of physical measurements to insure that the tomographic theoretical models are valid and that the coding accurately describes these models. Simulation may also be used to determine the detection limits of the reconstruction methodology. A description of the system, the theory, and a demonstration of the systems capabilities is provided in the paper. The hardware for this system is currently under development

  1. Robotics/Automated Systems Task Analysis and Description of Required Job Competencies Report. Task Analysis and Description of Required Job Competencies of Robotics/Automated Systems Technicians.

    Science.gov (United States)

    Hull, Daniel M.; Lovett, James E.

    This task analysis report for the Robotics/Automated Systems Technician (RAST) curriculum project first provides a RAST job description. It then discusses the task analysis, including the identification of tasks, the grouping of tasks according to major areas of specialty, and the comparison of the competencies to existing or new courses to…

  2. Description of the ternary system Cu-Ge-Te

    International Nuclear Information System (INIS)

    Dogguy, M.; Carcaly, C.; Rivet, J.; Flahaut, J.

    1977-01-01

    The Cu-Ge-Te ternary system has been studied by DTA and by crystallographic and metallographic analysis. The existence of a ternary compound Cu 2 GeTe 3 is demonstrated; this compound has a ternary incongruent melting point at 500 0 C. This ternary compound has a superstructure of a zinc blende type. The study shows the existence of five ternary eutectics. Two liquid-liquid miscibility gaps exist: the first is situated entirely in the ternary system; the second gives a monotectic region within the ternary system. (Auth.)

  3. Piping systems, containment pre-stressing and steel ventilation chimney

    International Nuclear Information System (INIS)

    Stuessi, U.

    1996-01-01

    Units 5 and 6 of NPP Kozloduy have been designed initially for seismic levels which are considered too low today. In the frame of an IAEA Coordinated Research Programme, a Swiss team has been commissioned by Natsionalna Elektricheska Kompania, Sofia, to analyse the relevant piping system, the containment prestressing and the steel ventilation chimney and to recommend upgrade measures for adequate seismic capacity where applicable. Seismic input had been specified by and agreed upon earlier by IAEA experts. The necessary investigations have been performed in 1995 and discussed with internationally recognized experts. The main results may be summarized as follows: Upgrades are necessary at different piping sy ports (additional snubbers or viscous dampers). These fixes can be done easily at low cost. The containment prestressing tendons are adequately designed for the specified load combinations. However, unfavourable construction features endanger the reliability. It is therefore strongly recommended to replace the tendons stepwise and to upgrade the existing monitoring system. Finally, the steel ventilation chimney may not withstand a seismic event, however the containment and diesel generator building will not be destroyed at possible impact by the chimney. On the other hand the roof of the main building has to be reinforced partially. It is recommended to continue the project for 1996 and 1997 to implement the upgrade measures mentioned above, to analyse the remaining piping systems and to consolidate all results obtained by different research groups of the IAEA programme with respect to piping systems including components and tanks

  4. Nuclear power safety reporting system feasibility analysis and concept description

    International Nuclear Information System (INIS)

    Finlayson, F.C.; Ims, J.R.; Hussman, T.A.

    1984-01-01

    The Aerospace Corporation is assisting the US Nuclear Regulatory Commission (NRC) in the evaluation of the potential attributes of a voluntary, nonpunitive data gathering system for identifying and quantifying the factors that contribute to the occurrence of significant safety problems involving humans in nuclear power plants. The objectives of the Aerospace Administration (FAA)/National Aeronautics and Space Administration (NASA) Aviation Safety Reporting System (ASRS) in order to determine whether it would be feasible to apply part (or all) of the ASRS concepts for collecting data on human factor related incidents to the nuclear industry; and (2) to identify and define the basic elements and requirements of a Nuclear Power Safety Reporting System (NPSRS), assuming the feasibility of implementing such a system was established

  5. Lagoa Real design. Description and evaluation of sampling system

    International Nuclear Information System (INIS)

    Hashizume, B.K.

    1982-10-01

    This report describes the samples preparation system of drilling from Lagoa Real Design, aiming obtainment representative fraction of the half from drilling outlier. The error of sampling + analysis and analytical accuracy was obtainment by delayed neutron analysis. (author)

  6. The vehicle data translator V3.0 system description.

    Science.gov (United States)

    2011-05-30

    With funding and support from the USDOT RITA and direction from the FHWA Road Weather Management Program, NCAR is developing a Vehicle Data Translator (VDT) software system that incorporates vehicle-based measurements of the road and surrounding atmo...

  7. Secure Automated Fabrication: a system design description (SDD), section 1

    International Nuclear Information System (INIS)

    Konze, G.M.; Thompson, M.L.; Wadekamper, D.C.; Zimmer, J.J.

    Information is presented concerning the conversion system to convert purified mixed nitrate solution to MO/sub x/ powder; powder preparation and pellet fabrication; sintering and pin loading; assembly fabrication; and scrap recovery

  8. W-314, waste transfer alternative piping system description

    International Nuclear Information System (INIS)

    Papp, I.G.

    1998-01-01

    It is proposed that the reliability, operability, and flexibility of the Retrieval Transfer System be substantially upgraded by replacing the planned single in-farm pipeline from the AN-AY-AZ-(SY) Tank Farm Complex to the AP Farm with three parallel pipelines outside the tank farms. The proposed system provides simplified and redundant routes for the various transfer missions, and prevents the risk of transfer gridlock when the privatization effort swings into full operation

  9. W-314, waste transfer alternative piping system description

    Energy Technology Data Exchange (ETDEWEB)

    Papp, I.G.

    1998-04-30

    It is proposed that the reliability, operability, and flexibility of the Retrieval Transfer System be substantially upgraded by replacing the planned single in-farm pipeline from the AN-AY-AZ-(SY) Tank Farm Complex to the AP Farm with three parallel pipelines outside the tank farms. The proposed system provides simplified and redundant routes for the various transfer missions, and prevents the risk of transfer gridlock when the privatization effort swings into full operation.

  10. Advanced Transport Operating System (ATOPS) utility library software description

    Science.gov (United States)

    Clinedinst, Winston C.; Slominski, Christopher J.; Dickson, Richard W.; Wolverton, David A.

    1993-01-01

    The individual software processes used in the flight computers on-board the Advanced Transport Operating System (ATOPS) aircraft have many common functional elements. A library of commonly used software modules was created for general uses among the processes. The library includes modules for mathematical computations, data formatting, system database interfacing, and condition handling. The modules available in the library and their associated calling requirements are described.

  11. Software Test Description (STD) for the Globally Relocatable Navy Tide/Atmospheric Modeling System (PCTides)

    National Research Council Canada - National Science Library

    Posey, Pamela

    2002-01-01

    The purpose of this Software Test Description (STD) is to establish formal test cases to be used by personnel tasked with the installation and verification of the Globally Relocatable Navy Tide/Atmospheric Modeling System (PCTides...

  12. Data Description Exchange Services for Heterogeneous Vehicle and Spaceport Control and Monitor Systems, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — CCT proposes an advanced data description exchange approach for space/spaceport systems that will provide a generic platform independent software capability for...

  13. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  14. Thermodynamic description of the Ta-W-Zr system

    International Nuclear Information System (INIS)

    Guo, Cuiping; Li, Changrong; Du, Zhenmin; Shang, Shunli

    2014-01-01

    The Ta-W, W-Zr and Ta-W-Zr systems are critically reviewed and modeled using the CALPHAD technique. The enthalpy of formation of the stoichiometric compound W 2 Zr in the W-Zr system is predicted from first-principles calculations. The solution phases (liquid, bcc and hcp) are modeled by the substitutional solution model. The compound W 2 Zr is treated with the formula (Ta,W) 2 Zr in the Ta-W-Zr system because of a significant solid solubility of Ta in W 2 Zr. All experimental data, including the Gibbs energy of formation, enthalpy of formation, activity of Ta and W of bcc phase at 1 200 K, Ta-W and W-Zr phase diagrams, and three isothermal sections of the Ta-W-Zr system at 1 073, 1 098, and 1 873 K, are reproduced in the present work. A set of self-consistent thermodynamic parameters of the Ta-W-Zr system is obtained.

  15. Thermodynamics of organic mixtures containing amines. VIII. Systems with quinoline

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, Juan Antonio [G.E.T.E.F., Grupo Especializado en Termodinamica de Equilibrio entre Fases, Departamento de Fisica Aplicada, Facultad de Ciencias, Universidad de Valladolid, E-47071 Valladolid (Spain)], E-mail: jagl@termo.uva.es; Domanska, Urszula; Zawadzki, Maciej [Physical Chemistry Division, Faculty of Chemistry, Warsaw University of Technology, 00-664 Warsaw (Poland)

    2008-08-15

    (Solid + liquid) equilibrium temperatures for mixtures containing quinoline and 1-dodecanol, 1-hexadecanol, or 1-octadecanol have been measured using a dynamic method. (Quinoline + benzene, +alkane, or +1-alkanol) systems were investigated using DISQUAC. The corresponding interaction parameters are reported. The model yields a good representation of molar excess Gibbs free energies, G{sup E}, molar excess enthalpies, H{sup E}, and of the (solid + liquid) equilibria, SLE. Interactional and structural effects were analysed comparing H{sup E} and the molar excess internal energy at constant volume, U{sub V}{sup E}. It was encountered that structural effects are very important in systems involving alkanes or 1-alkanols. Interactions between amine molecules are stronger in mixtures with quinoline than in those containing pyridine, which was ascribed to the higher polarizability of quinoline.

  16. Pressure suppression containment system for boiling water reactor

    Science.gov (United States)

    Gluntz, Douglas M.; Nesbitt, Loyd B.

    1997-01-01

    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  17. Description of an XRF system for multielemental analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wielopolski, L.; Zhang, R.; Cohn, S.H.

    1986-01-01

    An X-ray fluorescence (XRF) system which uses radioisotopes in an orthogonal configuration between the source, sample, and detector is described. The advantage of such a system is that for large (bulk) samples or for in vivo measurements the background due to Compton scattering in the sample is minimized. High reproducibility for nonuniform samples is obtained by reducing the sample size and thus the effects of non-uniformity in the spatial response of such a system. Germane to any accurate analytical method is the use of proper mathematical algorithms for data evaluation. The problem is acute, in particular, when photopeaks with low counting statistics are to be analyzed. In the case of a single photopeak on flat, background optimal energy window size, which maximizes the signal-to-noise ratio, for trapezoidal intergration is described. The sensitivity and minimum detection limit at different energies together with background considerations are discussed. 13 refs., 8 figs., 2 tabs.

  18. Description of an XRF system for multielemental analysis

    International Nuclear Information System (INIS)

    Wielopolski, L.; Zhang, R.; Cohn, S.H.

    1986-01-01

    An X-ray fluorescence (XRF) system which uses radioisotopes in an orthogonal configuration between the source, sample, and detector is described. The advantage of such a system is that for large (bulk) samples or for in vivo measurements the background due to Compton scattering in the sample is minimized. High reproducibility for nonuniform samples is obtained by reducing the sample size and thus the effects of non-uniformity in the spatial response of such a system. Germane to any accurate analytical method is the use of proper mathematical algorithms for data evaluation. The problem is acute, in particular, when photopeaks with low counting statistics are to be analyzed. In the case of a single photopeak on flat, background optimal energy window size, which maximizes the signal-to-noise ratio, for trapezoidal intergration is described. The sensitivity and minimum detection limit at different energies together with background considerations are discussed. 13 refs., 8 figs., 2 tabs

  19. Structure-rheology relations in sodium caseinate containing systems

    OpenAIRE

    Ruis, H.G.M.

    2007-01-01

    The general aim of the work described in this thesis was to investigate structure-rheologyrelations for dairy related products, focusing on model systems containing sodium caseinate. The acid inducedgelationof sodium caseinate, of sodium caseinate stabilized emulsions, and the effect of shear on the structure formation was characterized. Special attention was given to the sol-gel transition point, which was defined by a frequency independent loss tangent. It was shown that the sol-gel transit...

  20. Systems reliability Benchmark exercise part 1-Description and results

    International Nuclear Information System (INIS)

    Amendola, A.

    1986-01-01

    The report describes aims, rules and results of the Systems Reliability Benchmark Exercise, which has been performed in order to assess methods and procedures for reliability analysis of complex systems and involved a large number of European organizations active in NPP safety evaluation. The exercise included both qualitative and quantitative methods and was structured in such a way that separation of the effects of uncertainties in modelling and in data on the overall spread was made possible. Part I describes the way in which RBE has been performed, its main results and conclusions

  1. Physics Detector Simulation Facility Phase II system software description

    International Nuclear Information System (INIS)

    Scipioni, B.; Allen, J.; Chang, C.; Huang, J.; Liu, J.; Mestad, S.; Pan, J.; Marquez, M.; Estep, P.

    1993-05-01

    This paper presents the Physics Detector Simulation Facility (PDSF) Phase II system software. A key element in the design of a distributed computing environment for the PDSF has been the separation and distribution of the major functions. The facility has been designed to support batch and interactive processing, and to incorporate the file and tape storage systems. By distributing these functions, it is often possible to provide higher throughput and resource availability. Similarly, the design is intended to exploit event-level parallelism in an open distributed environment

  2. CMS Muon Alignment: System Description and first results

    CERN Document Server

    Sobron, M

    2008-01-01

    The CMS detector has been instrumented with a precise and complex opto-mechanical alignment subsystem that provides a common reference frame between Tracker and Muon detection systems by means of a net of laser beams. The system allows a continuous and accurate monitoring of the muon chambers positions with respect to the Tracker body. Preliminary results of operation during the test of the CMS 4T solenoid magnet, performed in 2006, are presented. These measurements complement the information provided by the use of survey techniques and the results of alignment algorithms based on muon tracks crossing the detector.

  3. An Automated Weather Research and Forecasting (WRF)-Based Nowcasting System: Software Description

    Science.gov (United States)

    2013-10-01

    14. ABSTRACT A Web service /Web interface software package has been engineered to address the need for an automated means to run the Weather Research...An Automated Weather Research and Forecasting (WRF)- Based Nowcasting System: Software Description by Stephen F. Kirby, Brian P. Reen, and...Based Nowcasting System: Software Description Stephen F. Kirby, Brian P. Reen, and Robert E. Dumais Jr. Computational and Information Sciences

  4. Secondary containment systems for bulk oil storage facilities

    International Nuclear Information System (INIS)

    Carr, B.A.

    1996-01-01

    The United States Environmental Protection Agency has conducted site inspections at several onshore bulk oil above ground storage facilities, to ensure that owners follow the spill prevention, control and countermeasure regulations. The four violations which were most frequently cited at these facilities were: (1) lack of a spill prevention plan, (2) lack of appropriate containment equipment to prevent discharged oil from reaching a navigable water course, (3) inadequate secondary containment structures, and (4) lack of an adequate quick drainage system in the facility tank loading/unloading area. Suggestions for feasible designs which would improve the impermeability of secondary containment for above ground storage tanks (AST) included the addition of a liner, retrofitting the bottom of an AST with a second steel plate, using a geosynthetic liner on top of the original bottom, installing a leak detection system in the interstitial space between the steel plates, or installing an under-tank liner with a leak detection system during construction of a new AST. 2 refs

  5. Description of Fracture Systems for External Criticality Reports

    International Nuclear Information System (INIS)

    Nicot, Jean-Philippe

    2001-01-01

    The purpose of this Analysis/Model Report (AMR) is to describe probabilistically the main features of the geometry of the fracture system in the vicinity of the repository. They will be used to determine the quantity of fissile material that could accumulate in the fractured rock underneath a waste package as it degrades. This AMR is to feed the geochemical calculations for external criticality reports. This AMR is done in accordance with the technical work plan (BSC (Bechtel SAIC Company) 2001 b). The scope of this AMR is restricted to the relevant parameters of the fracture system. The main parameters of interest are fracture aperture and fracture spacing distribution parameters. The relative orientation of the different fracture sets is also important because of its impact on criticality, but they will be set deterministically. The maximum accumulation of material depends primarily on the fracture porosity, combination of the fracture aperture, and fracture intensity. However, the fracture porosity itself is not sufficient to characterize the potential for accumulation of a fracture system. The fracture aperture is also important because it controls both the flow through the fracture and the potential plugging of the system. Other features contributing to the void space such as lithophysae are also investigated. On the other hand, no analysis of the matrix porosity is done. The parameters will be used in sensitivity analyses of geochemical calculations providing actinide accumulations and in the subsequent Monte Carlo criticality analyses

  6. Cold Vacuum Drying facility condensate collection system design description

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) condensate collection system (CCS). The function of the CCS is to collect cooling coil condensate from air-handling units in the CVDF and to isolate the condensate in collection tanks until the condensate is determined to be acceptable to drain to the effluent drain collection basin

  7. LLNL current meter array--concept and system description

    Energy Technology Data Exchange (ETDEWEB)

    Mantrom, D.D. [Lawrence Livermore National Lab., CA (United States)

    1994-11-15

    A measurement capability using a horizontal array of 10 S4 current meters mounted on a stiff floating structure with 35 m aperture has been developed to support interpretation of radar imaging of surface effects associated with internal waves. This system has been fielded three times and most recently, has collected data alongside the sea-surface footprint of a land-fixed radar imaging ship-generated internal waves. The underlying need for this measurement capability is described. The specifications resulting from this need are presented and the engineering design and deployment procedures of the platform and systems that resulted are described The current meter data are multiplexed along with meteorological and system status data on board the floating platform and are telemetered to a shore station and on to a data acquisition system. The raw data are recorded, and are then processed to form space-time images of current and strain rate (a spatial derivative of the current field). Examples of raw and processed data associated with ship-generated internal waves are presented.

  8. Operator symbols in the description of observable-state systems

    International Nuclear Information System (INIS)

    Lassner, G.A.

    1978-01-01

    For the observable-state system of finite degree of freedom N topological properties of the kernels and symbols belonging to the considered operators are investigated. For the operators of the observable algebra of rho + (delta) kernels and symbols are distributions and for density matrices p they are smooth functions

  9. Spontaneous nervous system concussion in dogs: A description of ...

    African Journals Online (AJOL)

    In human medicine, central nervous system (CNS) concussion is defined as a transient neurological dysfunction following a traumatic event, without evidence of structural abnormalities of the affected region on advanced diagnostic imaging. Depending on the anatomical region involved, three forms of concussive ...

  10. Description of the IV + V System Software Package.

    Science.gov (United States)

    Microcomputers for Information Management: An International Journal for Library and Information Services, 1984

    1984-01-01

    Describes the IV + V System, a software package designed by the Institut fur Maschinelle Dokumentation for the United Nations General Information Programme and UNISIST to support automation of local information and documentation services. Principle program features and functions outlined include input/output, databank, text image, output, and…

  11. Description and analysis of the RIES Internet voting system

    NARCIS (Netherlands)

    Hubbers, E.; Jacobs, B.P.F.; Schoenmakers, B.; Tilborg, van H.C.A.; Weger, de B.M.M.

    2008-01-01

    RIES is an evolving family of systems (RIES-2004, RIES-KOA, RIES-2008) for electronic elections via the Internet. It has been used in practice for medium scale elections for the Dutch District Water Control Boards and for expatriates in national parliament elections. We describe and analyze the

  12. FCA containment and surveillance (C/S) system

    International Nuclear Information System (INIS)

    Ogawa, Hironobu; Mukaiyama, Takehiko; Yokota, Yasuhiro.

    1994-11-01

    The Fast Critical Assembly (FCA) facility of the Japan Atomic Energy Research Inst. (JAERI) is internationally recognized as one of the most sensitive facility in the world from the viewpoint of international safeguards, because the facility possesses a large amount of metallic uranium and metallic plutonium, which are needed to perform various physical experiments. These material are subject to frequent verifications by the inspectorate, the International Atomic Energy Agency (IAEA) and the domestic authority (Science and Technology Agency of Japan, STA). Those verifications require inspectors to access to these materials for measurements and applications of seals. Human resources increase of irradiations and restrictions on the freedom of physical experiments, that are inevitably associated with these inspection activities, have been a serious problem that causes significant burdens for all relating parties. To decrease these burdens without any confliction with the inspection goals, an advanced comprehensive system of containment and surveillance has been developed. The FCA Containment and Surveillance (C/S) System consists of tow independent subsystems, i. e. Portal Monitor (P/M) and Penetration Monitor(PN/M). In this system the internal wall of the reactor building is used as a part of containment for the safeguards purpose, which enables the portal, that is installed at the internal wall of the reactor building, to be used as an area for monitoring of any removal of nuclear material. A metal detector of high sensitivity has been selected for the system since all nuclear materials possessed by the FCA has metallic forms. The internal wall has several penetrations for utility purposes, which should also be monitored for the purpose of detecting any removal of nuclear material from the reactor core area. A penetration monitor system has been developed for this purpose. This report describes functions of the system and their operation procedures. (author)

  13. Office of River Protection Integrated Safety Management System Description

    Energy Technology Data Exchange (ETDEWEB)

    CLARK, D.L.

    1999-08-09

    Revision O was never issued. Finding safe and environmentally sound methods of storage and disposal of 54 million gallons of highly radioactive waste contained in 177 underground tanks is the largest challenge of Hanford cleanup. TWRS was established in 1991 and continues to integrate all aspects of the treatment and management of the high-level radioactive waste tanks. In fiscal Year 1997, program objectives were advanced in a number of areas. RL TWRS refocused the program toward retrieving, treating, and immobilizing the tank wastes, while maintaining safety as first priority. Moving from a mode of storing the wastes to getting the waste out of the tanks will provide the greatest cleanup return on the investment and eliminate costly mortgage continuance. There were a number of safety-related achievements in FY1997. The first high priority safety issue was resolved with the removal of 16 tanks from the ''Wyden Watch List''. The list, brought forward by Senator Ron Wyden of Oregon, identified various Hanford safety issues needing attention. One of these issues was ferrocyanide, a chemical present in 24 tanks. Although ferrocyanide can ignite at high temperature, analysis found that the chemical has decomposed into harmless compounds and is no longer a concern.

  14. Disruption management for truck appointment system at a container terminal

    DEFF Research Database (Denmark)

    Li, N.; Chen, Gang; Jin, Z.

    2016-01-01

    -appointed arrivals at a container terminal that is running an appointment system. Second, we propose some response strategies to cope with different levels of disruptions, and evaluate their resilience ability with two Key Performance Indicators (KPIs): total waiting time of on-time trucks and total idling emissions...... of all trucks, in order to balance the service quality to punctual arrivals and green performance of the whole system. Third, we conduct a sensitivity analysis using a discrete event simulation to understand the performance of the proposed strategies. Considering both KPIs, the best strategy in most......-crane moving distance, especially when the first KPI is given lower weight than the second one....

  15. Cold Vacuum Drying facility effluent drain system design description

    International Nuclear Information System (INIS)

    PITKOFF, C.C.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) effluent drain system (EFS). The primary function of the EFS is to collect and transport fire suppression water discharged into a CVDF process bay to a retention basin located outside the facility. The EFS also provides confinement of spills that occur inside a process bay and allows non-contaminated water that drains to the process bay sumps to be collected until sampling and analysis are complete

  16. High-Level Waste System Process Interface Description

    International Nuclear Information System (INIS)

    D'Entremont, P.D.

    1999-01-01

    The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment

  17. Asymptotic expansion and statistical description of turbulent systems

    International Nuclear Information System (INIS)

    Hagan, W.K. III.

    1986-01-01

    A new approach to studying turbulent systems is presented in which an asymptotic expansion of the general dynamical equations is performed prior to the application of statistical methods for describing the evolution of the system. This approach has been applied to two specific systems: anomalous drift wave turbulence in plasmas and homogeneous, isotropic turbulence in fluids. For the plasma case, the time and length scales of the turbulent state result in the asymptotic expansion of the Vlasov/Poisson equations taking the form of nonlinear gyrokinetic theory. Questions regarding this theory and modern Hamiltonian perturbation methods are discussed and resolved. A new alternative Hamiltonian method is described. The Eulerian Direct Interaction Approximation (EDIA) is slightly reformulated and applied to the equations of nonlinear gyrokinetic theory. Using a similarity transformation technique, expressions for the thermal diffusivity are derived from the EDIA equations for various geometries, including a tokamak. In particular, the unique result for generalized geometry may be of use in evaluating fusion reactor designs and theories of anomalous thermal transport in tokamaks. Finally, a new and useful property of the EDIA is pointed out. For the fluid case, an asymptotic expansion is applied to the Navier-Stokes equation and the results lead to the speculation that such an approach may resolve the problem of predicting the Kolmogorov inertial range energy spectrum for homogeneous, isotropic turbulence. 45 refs., 3 figs

  18. System Advisor Model, SAM 2014.1.14: General Description

    Energy Technology Data Exchange (ETDEWEB)

    Blair, Nate [National Renewable Energy Lab. (NREL), Golden, CO (United States); Dobos, Aron P. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Freeman, Janine [National Renewable Energy Lab. (NREL), Golden, CO (United States); Neises, Ty [National Renewable Energy Lab. (NREL), Golden, CO (United States); Wagner, Michael [National Renewable Energy Lab. (NREL), Golden, CO (United States); Ferguson, Tom [Global Resources, Northbrook, IL (United States); Gilman, Paul [National Renewable Energy Lab. (NREL), Golden, CO (United States); Janzou, Steven [Janzou Consulting, Idaho Springs, CO (United States)

    2014-02-01

    This document describes the capabilities of the U.S. Department of Energy and National Renewable Energy Laboratory's System Advisor Model (SAM), Version 2013.9.20, released on September 9, 2013. SAM is a computer model that calculates performance and financial metrics of renewable energy systems. Project developers, policy makers, equipment manufacturers, and researchers use graphs and tables of SAM results in the process of evaluating financial, technology, and incentive options for renewable energy projects. SAM simulates the performance of photovoltaic, concentrating solar power, solar water heating, wind, geothermal, biomass, and conventional power systems. The financial model can represent financial structures for projects that either buy and sell electricity at retail rates (residential and commercial) or sell electricity at a price determined in a power purchase agreement (utility). SAM's advanced simulation options facilitate parametric and sensitivity analyses, and statistical analysis capabilities are available for Monte Carlo simulation and weather variability (P50/P90) studies. SAM can also read input variables from Microsoft Excel worksheets. For software developers, the SAM software development kit (SDK) makes it possible to use SAM simulation modules in their applications written in C/C++, C#, Java, Python, and MATLAB. NREL provides both SAM and the SDK as free downloads at http://sam.nrel.gov. Technical support and more information about the software are available on the website.

  19. Concept for a Satellite-Based Advanced Air Traffic Management System : Volume 4. Operational Description and Qualitative Assessment.

    Science.gov (United States)

    1974-02-01

    The volume presents a description of how the Satellite-Based Advanced Air Traffic Management System (SAATMS) operates and a qualitative assessment of the system. The operational description includes the services, functions, and tasks performed by the...

  20. Blind Decoding of Multiple Description Codes over OFDM Systems via Sequential Monte Carlo

    Directory of Open Access Journals (Sweden)

    Guo Dong

    2005-01-01

    Full Text Available We consider the problem of transmitting a continuous source through an OFDM system. Multiple description scalar quantization (MDSQ is applied to the source signal, resulting in two correlated source descriptions. The two descriptions are then OFDM modulated and transmitted through two parallel frequency-selective fading channels. At the receiver, a blind turbo receiver is developed for joint OFDM demodulation and MDSQ decoding. Transformation of the extrinsic information of the two descriptions are exchanged between each other to improve system performance. A blind soft-input soft-output OFDM detector is developed, which is based on the techniques of importance sampling and resampling. Such a detector is capable of exchanging the so-called extrinsic information with the other component in the above turbo receiver, and successively improving the overall receiver performance. Finally, we also treat channel-coded systems, and a novel blind turbo receiver is developed for joint demodulation, channel decoding, and MDSQ source decoding.

  1. Shared Task System Description: Frustratingly Hard Compositionality Prediction

    DEFF Research Database (Denmark)

    Johannsen, Anders Trærup; Martinez Alonso, Hector; Rishøj, Christian

    2011-01-01

    , and the likelihood of long translation equivalents in other languages. Many of the features we considered correlated significantly with human compositionality scores, but in support vector regression experiments we obtained the best results using only COALS-based endocentricity scores. Our system was nevertheless......We considered a wide range of features for the DiSCo 2011 shared task about compositionality prediction for word pairs, including COALS-based endocentricity scores, compositionality scores based on distributional clusters, statistics about wordnet-induced paraphrases, hyphenation...

  2. Containment hydrogen removal system for a nuclear power plant

    International Nuclear Information System (INIS)

    Callaghan, V.M.; Flynn, E.P.; Pokora, B.M.

    1984-01-01

    A hydrogen removal system (10) separates hydrogen from the containment atmosphere of a nuclear power plant using a hydrogen permeable membrane separator (30). Water vapor is removed by condenser (14) from a gas stream withdrawn from the containment atmosphere. The gas stream is then compressed by compressor (24) and cooled (28,34) to the operating temperature of the hydrogen permeable membrane separator (30). The separator (30) separates the gas stream into a first stream, rich in hydrogen permeate, and a second stream that is hydrogen depleted. The separated hydrogen is passed through a charcoal adsorber (48) to adsorb radioactive particles that have passed through the hydrogen permeable membrane (44). The hydrogen is then flared in gas burner (52) with atmospheric air and the combustion products vented to the plant vent. The hydrogen depleted stream is returned to containment through a regenerative heat exchanger (28) and expander (60). Energy is extracted from the expander (60) to drive the compressor (24) thereby reducing the energy input necessary to drive the compressor (24) and thus reducing the hydrogen removal system (10) power requirements

  3. Description of the National Hydrologic Model for use with the Precipitation-Runoff Modeling System (PRMS)

    Science.gov (United States)

    Regan, R. Steven; Markstrom, Steven L.; Hay, Lauren E.; Viger, Roland J.; Norton, Parker A.; Driscoll, Jessica M.; LaFontaine, Jacob H.

    2018-01-08

    This report documents several components of the U.S. Geological Survey National Hydrologic Model of the conterminous United States for use with the Precipitation-Runoff Modeling System (PRMS). It provides descriptions of the (1) National Hydrologic Model, (2) Geospatial Fabric for National Hydrologic Modeling, (3) PRMS hydrologic simulation code, (4) parameters and estimation methods used to compute spatially and temporally distributed default values as required by PRMS, (5) National Hydrologic Model Parameter Database, and (6) model extraction tool named Bandit. The National Hydrologic Model Parameter Database contains values for all PRMS parameters used in the National Hydrologic Model. The methods and national datasets used to estimate all the PRMS parameters are described. Some parameter values are derived from characteristics of topography, land cover, soils, geology, and hydrography using traditional Geographic Information System methods. Other parameters are set to long-established default values and computation of initial values. Additionally, methods (statistical, sensitivity, calibration, and algebraic) were developed to compute parameter values on the basis of a variety of nationally-consistent datasets. Values in the National Hydrologic Model Parameter Database can periodically be updated on the basis of new parameter estimation methods and as additional national datasets become available. A companion ScienceBase resource provides a set of static parameter values as well as images of spatially-distributed parameters associated with PRMS states and fluxes for each Hydrologic Response Unit across the conterminuous United States.

  4. A description of smallholder pig production systems in eastern Indonesia.

    Science.gov (United States)

    Leslie, Edwina E C; Geong, Maria; Abdurrahman, Muktasam; Ward, Michael P; Toribio, Jenny-Ann L M L

    2015-03-01

    Pig farming is a common practice among smallholder farmers in Nusa Tenggara Timur province (NTT), eastern Indonesia. To understand their production systems a survey of smallholder pig farmers was conducted. Eighteen villages were randomly selected across West Timor, Flores and Sumba islands, and 289 pig farmers were interviewed. Information on pig management, biosecurity practices, pig movements and knowledge of pig health and disease, specifically classical swine fever was collected. The mean number of pigs per herd was 5.0 (not including piglets), and total marketable herd size (pigs≥two months of age) did not differ significantly between islands (P=0.215). Chickens (71%) and dogs (62%) were the most commonly kept animal species in addition to pigs. Pigs were mainly kept as a secondary income source (69%) and 83% of farmers owned at least one sow. Seventy-four percent (74%) of pigs were housed in a kandang (small bamboo pen) and 25% were tethered. Pig feeds were primarily locally sourced agricultural products (93%). The majority of farmers had no knowledge of classical swine fever (91%) and biosecurity practices were minimal. Forty-five percent (45%) reported to consuming a pig when it died and 74% failed to report cases of sick or dead pigs to appropriate authorities. Sixty-five percent (65%) of farmers reported that a veterinarian or animal health worker had never visited their village. Backyard slaughter was common practice (55%), with meat mainly used for home consumption (89%). Most (73%) farmers purchased pigs in order to raise the animal on their farm with 36% purchasing at least one pig within the last year. Predominantly fattener pigs (34%) were given as gifts for celebratory events, most commonly for funerals (32%), traditional ceremonies (27%) and marriages (10%). For improved productivity of this traditional low-input system, research incorporating farming training and improved knowledge on pig disease and biosecurity needs to be integrated with

  5. Secondary containment system for a high tritium research cryostat

    International Nuclear Information System (INIS)

    Tsugawa, R.T.; Fearon, D.; Souers, P.C.; Hickman, R.G.; Roberts, P.E.

    1976-01-01

    A 4.2- to 300-K liquid helium cryostat has been constructed for cryogenic samples of D--T containing up to 4 x 10 14 dis/s (10,000 Ci) of tritium radioactivity. The cryostat is enclosed in a secondary box, which acts as the ultimate container in case of a tritium release. Dry argon is flushed through the box, and the box atmosphere is monitored for tritium, oxygen, and water vapor. A rupture disk and abort tank protect the box atmosphere in case the sample cell breaks. If tritium breaks into the box, a powdered uranium getter trap reduces the 4 x 10 14 dis/s (10,000 Ci) to 4 x 10 9 dis/s (0.1 Ci) in 24 h. A backup palladium--zeolite getter system goes into operation if an overabundance of oxygen contaminates the uranium getter

  6. 40 CFR 90.421 - Dilute gaseous exhaust sampling and analytical system description.

    Science.gov (United States)

    2010-07-01

    ... gas mixture temperature, measured at a point immediately ahead of the critical flow venturi, must be... analytical system description. (a) General. The exhaust gas sampling system described in this section is... requirements are as follows: (1) This sampling system requires the use of a Positive Displacement Pump—Constant...

  7. Description and evaluation of the CASA dual-Doppler system

    Science.gov (United States)

    Martinez, Matthew

    2011-12-01

    Long range weather surveillance radars are designed for observing weather events for hundreds of kilometers from the radar and operate over a large coverage domain independently of weather conditions. As a result a loss in spatial resolution and limited temporal sampling of the weather phenomenon occurs. Due to the curvature of the Earth, long-range weather radars tend to make the majority of their precipitation and wind observations in the middle to upper troposphere, resulting in missed features associates with severe weather occurring in the lowest three kilometers of the troposphere. The spacing of long-range weather radars in the United States limits the feasibility of using dual-Doppler wind retrievals that would provide valuable information on the kinematics of weather events to end-users and researchers. The National Science Foundation Center for Collaborative Adapting Sensing of the Atmosphere (CASA) aims to change the current weather sensing model by increasing coverage of the lowest three kilometers of the troposphere by using densely spaced networked short-range weather radars. CASA has deployed a network of these radars in south-western Oklahoma, known as Integrated Project 1 (IP1). The individual radars are adaptively steered by an automated system known as the Meteorological Command and Control (MCC). The geometry of the IP1 network is such that the coverage domains of the individual radars are overlapping. A dual-Doppler system has been developed for the IP1 network which takes advantage of the overlapping coverage domains. The system is comprised of two subsystems, scan optimization and wind field retrieval. The scan strategy subsystem uses the DCAS model and the number of dual-Doppler pairs in the IP1 network to minimizes the normalized standard deviation in the wind field retrieval. The scan strategy subsystem also minimizes the synchronization error between two radars. The retrieval itself is comprised of two steps, data resampling and the

  8. ITER hydrogen isotope separation system conceptual design description

    International Nuclear Information System (INIS)

    Busigin, A.; Sood, S.K.; Kveton, O.K.; Dinner, P.J.; Murdoch, D.K.; Leger, D.

    1990-01-01

    This paper presents integrated hydrogen Isotope Separation System (ISS) designs for ITER based on requirements for plasma exhaust processing, neutral beam injection deuterium cleanup, pellet injector propellant detritiation, waste water detritiation, and breeding blanket detritiation. Specific ISS designs are developed for a machine with an aqueous lithium salt blanket (ALSB) and a machine with a solid ceramic breeding blanket (SBB). The differences in the ISS designs arising from the different blanket concepts are highlighted. It is found that the ISS designs for the two blanket concepts considered are very similar with the only major difference being the requirement for an additional large water distillation column for ALSB water detritiation. The extraction of tritium from the ALSB is based on flash evaporation to separate the blanket water from the dissolved Li salt, with the tritiated water then being fed to the ISS for detritiation. This technology is considered to be relatively well understood in comparison to front-end processes for SBB detritiation. In the design of the cryogenic distillation portion of the ISS, it was found that the tritium inventory could be very large (> 600 g) unless specific design measures were taken to reduce it. In the designs which are presented, the tritium inventory has been reduced to about 180 g, which is less than the ITER single-failure release limit of 200 g. Further design optimization and isolation of components is expected to reduce the inventory further. (orig.)

  9. Description of waste pretreatment and interfacing systems dynamic simulation model

    International Nuclear Information System (INIS)

    Garbrick, D.J.; Zimmerman, B.D.

    1995-05-01

    The Waste Pretreatment and Interfacing Systems Dynamic Simulation Model was created to investigate the required pretreatment facility processing rates for both high level and low level waste so that the vitrification of tank waste can be completed according to the milestones defined in the Tri-Party Agreement (TPA). In order to achieve this objective, the processes upstream and downstream of the pretreatment facilities must also be included. The simulation model starts with retrieval of tank waste and ends with vitrification for both low level and high level wastes. This report describes the results of three simulation cases: one based on suggested average facility processing rates, one with facility rates determined so that approximately 6 new DSTs are required, and one with facility rates determined so that approximately no new DSTs are required. It appears, based on the simulation results, that reasonable facility processing rates can be selected so that no new DSTs are required by the TWRS program. However, this conclusion must be viewed with respect to the modeling assumptions, described in detail in the report. Also included in the report, in an appendix, are results of two sensitivity cases: one with glass plant water recycle steams recycled versus not recycled, and one employing the TPA SST retrieval schedule versus a more uniform SST retrieval schedule. Both recycling and retrieval schedule appear to have a significant impact on overall tank usage

  10. Simulation of Molecular Transport in Systems Containing Mobile Obstacles.

    Science.gov (United States)

    Polanowski, Piotr; Sikorski, Andrzej

    2016-08-04

    In this paper, we investigate the movement of molecules in crowded environments with obstacles undergoing Brownian motion by means of extensive Monte Carlo simulations. Our investigations were performed using the dynamic lattice liquid model, which was based on the cooperative movement concept and allowed to mimic systems at high densities where the motion of all elements (obstacles as well as moving particles) were highly correlated. The crowded environments are modeled on a two-dimensional triangular lattice containing obstacles (particles whose mobility was significantly reduced) moving by a Brownian motion. The subdiffusive motion of both elements in the system was analyzed. It was shown that the percolation transition does not exist in such systems in spite of the cooperative character of the particles' motion. The reduction of the obstacle mobility leads to the longer caging of liquid particles by mobile obstacles.

  11. Degradation and failure characteristics of NPP containment protective coating systems

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.

    2000-03-30

    A research program to investigate the performance and potential for failure of Service Level 1 coating systems used in nuclear power plant containment is in progress. The research activities are aligned to address phenomena important to cause failure as identified by the industry coatings expert panel. The period of interest for performance covers the time from application of the coating through 40 years of service, followed by a medium-to-large break loss-of-coolant accident scenario, which is a design basis accident (DBA) scenario. The interactive program elements are discussed in this report and the application of these elements to the System 5 coating system (polyamide epoxy primer, carbon steel substrate) is used to evaluate performance.

  12. Degradation and failure characteristics of NPP containment protective coating systems

    International Nuclear Information System (INIS)

    Sindelar, R.L.

    2000-01-01

    A research program to investigate the performance and potential for failure of Service Level 1 coating systems used in nuclear power plant containment is in progress. The research activities are aligned to address phenomena important to cause failure as identified by the industry coatings expert panel. The period of interest for performance covers the time from application of the coating through 40 years of service, followed by a medium-to-large break loss-of-coolant accident scenario, which is a design basis accident (DBA) scenario. The interactive program elements are discussed in this report and the application of these elements to the System 5 coating system (polyamide epoxy primer, carbon steel substrate) is used to evaluate performance

  13. Design compliance matrix waste sample container filling system for nested, fixed-depth sampling system

    International Nuclear Information System (INIS)

    BOGER, R.M.

    1999-01-01

    This design compliance matrix document provides specific design related functional characteristics, constraints, and requirements for the container filling system that is part of the nested, fixed-depth sampling system. This document addresses performance, external interfaces, ALARA, Authorization Basis, environmental and design code requirements for the container filling system. The container filling system will interface with the waste stream from the fluidic pumping channels of the nested, fixed-depth sampling system and will fill containers with waste that meet the Resource Conservation and Recovery Act (RCRA) criteria for waste that contains volatile and semi-volatile organic materials. The specifications for the nested, fixed-depth sampling system are described in a Level 2 Specification document (HNF-3483, Rev. 1). The basis for this design compliance matrix document is the Tank Waste Remediation System (TWRS) desk instructions for design Compliance matrix documents (PI-CP-008-00, Rev. 0)

  14. Decision support system for containment and release management

    Energy Technology Data Exchange (ETDEWEB)

    Oosterhuis, B [Twente Univ., Enschede (Netherlands). Computer Science Dept.

    1995-09-01

    The Containment and Release Management project was carried out within the Reinforced Concerted Action Programme for Accident Management Support and partly financed by the European Union. In this report a prototype of an accident management support system is presented. The support system integrates several concepts from accident management research, like safety objective trees, severe accident phenomena, calculation models and an emergency response data system. These concepts are provided by the prototype in such a way that the decision making process of accident management is supported. The prototype application is demonstrated by process data taken from a severe accident scenario for a pressurized water reactor (PWR) that was simulated with the thermohydraulic computer program MAAP. The prototype was derived from a decision support framework based on a decision theory. For established and innovative concepts from accident management research it is pointed out in which way these concepts can support accident management and how these concepts can be integrated. This approach is generic in two ways; it applies to both pressurized and boiling water reactors and it applies to both in vessel management and containment and release management. The prototype application was developed in Multimedia Toolbox 3.0 and requires at least a 386 PC with 4 MB memory, 6 MB free disk space and MS Windows 3.1. (orig.).

  15. Decision support system for containment and release management

    International Nuclear Information System (INIS)

    Oosterhuis, B.

    1995-09-01

    The Containment and Release Management project was carried out within the Reinforced Concerted Action Programme for Accident Management Support and partly financed by the European Union. In this report a prototype of an accident management support system is presented. The support system integrates several concepts from accident management research, like safety objective trees, severe accident phenomena, calculation models and an emergency response data system. These concepts are provided by the prototype in such a way that the decision making process of accident management is supported. The prototype application is demonstrated by process data taken from a severe accident scenario for a pressurized water reactor (PWR) that was simulated with the thermohydraulic computer program MAAP. The prototype was derived from a decision support framework based on a decision theory. For established and innovative concepts from accident management research it is pointed out in which way these concepts can support accident management and how these concepts can be integrated. This approach is generic in two ways; it applies to both pressurized and boiling water reactors and it applies to both in vessel management and containment and release management. The prototype application was developed in Multimedia Toolbox 3.0 and requires at least a 386 PC with 4 MB memory, 6 MB free disk space and MS Windows 3.1. (orig.)

  16. SLUDGE TREATMENT PROJECT ENGINEERED CONTAINER RETRIEVAL AND TRANSFER SYSTEM PRELMINARY DESIGN HAZARD AND OPERABILITY STUDY

    Energy Technology Data Exchange (ETDEWEB)

    CARRO CA

    2011-07-15

    This Hazard and Operability (HAZOP) study addresses the Sludge Treatment Project (STP) Engineered Container Retrieval and Transfer System (ECRTS) preliminary design for retrieving sludge from underwater engineered containers located in the 105-K West (KW) Basin, transferring the sludge as a sludge-water slurry (hereafter referred to as 'slurry') to a Sludge Transport and Storage Container (STSC) located in a Modified KW Basin Annex, and preparing the STSC for transport to T Plant using the Sludge Transport System (STS). There are six, underwater engineered containers located in the KW Basin that, at the time of sludge retrieval, will contain an estimated volume of 5.2 m{sup 3} of KW Basin floor and pit sludge, 18.4 m{sup 3} of 105-K East (KE) Basin floor, pit, and canister sludge, and 3.5 m{sup 3} of settler tank sludge. The KE and KW Basin sludge consists of fuel corrosion products (including metallic uranium, and fission and activation products), small fuel fragments, iron and aluminum oxide, sand, dirt, operational debris, and biological debris. The settler tank sludge consists of sludge generated by the washing of KE and KW Basin fuel in the Primary Clean Machine. A detailed description of the origin of sludge and its chemical and physical characteristics can be found in HNF-41051, Preliminary STP Container and Settler Sludge Process System Description and Material Balance. In summary, the ECRTS retrieves sludge from the engineered containers and hydraulically transfers it as a slurry into an STSC positioned within a trailer-mounted STS cask located in a Modified KW Basin Annex. The slurry is allowed to settle within the STSC to concentrate the solids and clarify the supernate. After a prescribed settling period the supernate is decanted. The decanted supernate is filtered through a sand filter and returned to the basin. Subsequent batches of slurry are added to the STSC, settled, and excess supernate removed until the prescribed quantity of sludge is

  17. Sludge Treatment Project Engineered Container Retrieval And Transfer System Prelminary Design Hazard And Operability Study

    International Nuclear Information System (INIS)

    Carro, C.A.

    2011-01-01

    This Hazard and Operability (HAZOP) study addresses the Sludge Treatment Project (STP) Engineered Container Retrieval and Transfer System (ECRTS) preliminary design for retrieving sludge from underwater engineered containers located in the 105-K West (KW) Basin, transferring the sludge as a sludge-water slurry (hereafter referred to as 'slurry') to a Sludge Transport and Storage Container (STSC) located in a Modified KW Basin Annex, and preparing the STSC for transport to T Plant using the Sludge Transport System (STS). There are six, underwater engineered containers located in the KW Basin that, at the time of sludge retrieval, will contain an estimated volume of 5.2 m 3 of KW Basin floor and pit sludge, 18.4 m 3 of 105-K East (KE) Basin floor, pit, and canister sludge, and 3.5 m 3 of settler tank sludge. The KE and KW Basin sludge consists of fuel corrosion products (including metallic uranium, and fission and activation products), small fuel fragments, iron and aluminum oxide, sand, dirt, operational debris, and biological debris. The settler tank sludge consists of sludge generated by the washing of KE and KW Basin fuel in the Primary Clean Machine. A detailed description of the origin of sludge and its chemical and physical characteristics can be found in HNF-41051, Preliminary STP Container and Settler Sludge Process System Description and Material Balance. In summary, the ECRTS retrieves sludge from the engineered containers and hydraulically transfers it as a slurry into an STSC positioned within a trailer-mounted STS cask located in a Modified KW Basin Annex. The slurry is allowed to settle within the STSC to concentrate the solids and clarify the supernate. After a prescribed settling period the supernate is decanted. The decanted supernate is filtered through a sand filter and returned to the basin. Subsequent batches of slurry are added to the STSC, settled, and excess supernate removed until the prescribed quantity of sludge is collected. The sand

  18. Computer system design description for SY-101 hydrogen mitigation test project data acquisition and control system (DACS-1)

    International Nuclear Information System (INIS)

    Ermi, A.M.

    1997-01-01

    Description of the Proposed Activity/REPORTABLE OCCURRENCE or PIAB: This ECN changes the computer systems design description support document describing the computers system used to control, monitor and archive the processes and outputs associated with the Hydrogen Mitigation Test Pump installed in SY-101. There is no new activity or procedure associated with the updating of this reference document. The updating of this computer system design description maintains an agreed upon documentation program initiated within the test program and carried into operations at time of turnover to maintain configuration control as outlined by design authority practicing guidelines. There are no new credible failure modes associated with the updating of information in a support description document. The failure analysis of each change was reviewed at the time of implementation of the Systems Change Request for all the processes changed. This document simply provides a history of implementation and current system status

  19. Westinghouse containment filtered venting system wet scrubber technology

    International Nuclear Information System (INIS)

    Kristensson, S.; Nilsson, P-O.

    2014-01-01

    Following the Fukushima event Westinghouse has further developed and enhanced its filtered containment venting system (FCVS) product line. The filtration efficiency of the proven FILTRA-MVSS system installed at all Swedish NPPs as well as at the Muhelberg plant in Switzerland has been enhanced and a new wet scrubber design, SVEN (Safety Venting), based on the FILTRA-MVSS tradition, developed. To meet increased filtration requirements for organic iodine these two wet scrubber products have been complemented with a zeolite module. The offering of a select choice of products allows for a better adjustment to the specific constraints and needs of each nuclear power station that is planning for the installation of such a system. The FILTRA-MVSS (MVSS=Multi Venturi Scrubber System) is a wet containment filtered vent system that uses multiple venturies to create an interaction between the vent gases and the scrubber media allowing for removal of aerosols and gaseous iodines in a very efficient manner. The FILTRA-MVSS was originally developed to meet stringent requirements on autonomy and maintained filtration efficiency over a wide range of venting conditions. The system was jointly developed in the late 80's by ABB Atom and ABB Flaekt, today Westinghouse and Alstom. Following installations in Sweden and Switzerland the system was further developed by replacement of the gravel-bed moisture separator with a standard demister and by addition of a set of sintered metal fibre filter cartridges placed after the moisture separator step. The system is today offered as a modular steel tank design to simplify installation at site. To reduce complexity and delivery time Westinghouse has developed an alternative design in which the venturi module is replaced by a submerged metal fibre filter cartridges module. This new wet scrubber design, SVEN (patent pending), provides a flexible, compact, and lower weight system, while still preserving and even enhancing the filtration

  20. A remote inspection system for use inside reactor containment vessels

    International Nuclear Information System (INIS)

    Aoki, Toshihiko; Kashiwai, Jun-ichi; Yamamoto, Ikuo; Fukada, Koichi; Yamanaka, Yoshinobu.

    1985-01-01

    The harsh environment in the reactor-containment vesels of pressurized-water reactor nuclear-power plants precludes the possibility of direct circuit inspection; a remote-inspection system is essential. A robot for performing this task must not only be able to withstand the harsh conditions but must also be small and maneuverable enough to function effectively within complex and confined spaces. The article describes a monorail-type remote-inspection robot developed by Mitsubishi Electric to meet these needs, which is now under trial production and testing. (author)

  1. A stochastic killing system for biological containment of Escherichia coli

    DEFF Research Database (Denmark)

    Klemm, P.; Jensen, Lars Bogø; Molin, Søren

    1995-01-01

    Bacteria with a stochastic conditional lethal containment system have been constructed. The invertible switch promoter located upstream of the fimA gene from Escherichia coli was inserted as expression cassette in front of the Lethal gef gene deleted of its own natural promoter. The resulting...... fusion was placed on a plasmid and transformed to E. coli. The phenotype connected with the presence of such a plasmid was to reduce the population growth rate with increasing significance as the cell growth rate was reduced. In very fast growing cells, there was no measurable effect on growth rate. When...

  2. Container lid gasket protective strip for double door transfer system

    Science.gov (United States)

    Allen, Jr., Burgess M

    2013-02-19

    An apparatus and a process for forming a protective barrier seal along a "ring of concern" of a transfer container used with double door systems is provided. A protective substrate is supplied between a "ring of concern" and a safety cover in which an adhesive layer of the substrate engages the "ring of concern". A compressive foam strip along an opposite side of the substrate engages a safety cover such that a compressive force is maintained between the "ring of concern" and the adhesive layer of the substrate.

  3. System for indicating the level of material in a container

    International Nuclear Information System (INIS)

    Erb, T.L.

    1980-01-01

    In a radiation detecting system for controlling the level of material in a container, the first counter accumulates pulses generated by a geiger tube at a rate related to the level of material and a second counter accumulates clock pulses. A race condition is established between a NAND circuit indicating that the first counter has reached a predetermined total, and a NAND circuit indicating that the second counter has reached a second predetermined total representing a fixed counting interval. The first NAND circuit to respond to its predetermined total actuates a circuit to reset both counters and, if indicative of the material level being below a predetermined minimum, actuates an alarm or operates a control circuit to add material to the container. In the example shown, an additional NAND circuit responds to a different count in the first counter which count in the same time interval corresponds to a higher level, and when material is being added to the container, the race condition is between two NAND circuits. The effect of this is to provide a hysteresis effect preventing the circuit from 'hunting' around one level of material. (author)

  4. Cold Vacuum Drying (CVD) Facility Vacuum Purge System Chilled Water System Design Description. System 47-4

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    2000-01-01

    This system design description (SDD) addresses the Vacuum Purge System Chilled Water (VPSCHW) system. The discussion that follows is limited to the VPSCHW system and its interfaces with associated systems. The reader's attention is directed to Drawings H-1-82162, Cold Vacuum Drying Facility Process Equipment Skid PandID Vacuum System, and H-1-82224, Cold Vacuum Drying Facility Mechanical Utilities Process Chilled Water PandID. Figure 1-1 shows the location and equipment arrangement for the VPSCHW system. The VPSCHW system provides chilled water to the Vacuum Purge System (VPS). The chilled water provides the ability to condense water from the multi-canister overpack (MCO) outlet gases during the MCO vacuum and purge cycles. By condensing water from the MCO purge gas, the VPS can assist in drying the contents of the MCO

  5. Design of Training Systems, Phase II Report, Volume III; Model Program Descriptions and Operating Procedures. TAEG Report No. 12-2.

    Science.gov (United States)

    Naval Training Equipment Center, Orlando, FL. Training Analysis and Evaluation Group.

    The Design of Training Systems (DOTS) project was initiated by the Department of Defense (DOD) to develop tools for the effective management of military training organizations. Volume 3 contains the model and data base program descriptions and operating procedures designed for phase 2 of the project. Flow charts and program listings for the…

  6. A consistent description of kinetics and hydrodynamics of quantum Bose-systems

    Directory of Open Access Journals (Sweden)

    P.A.Hlushak

    2004-01-01

    Full Text Available A consistent approach to the description of kinetics and hydrodynamics of many-Boson systems is proposed. The generalized transport equations for strongly and weakly nonequilibrium Bose systems are obtained. Here we use the method of nonequilibrium statistical operator by D.N. Zubarev. New equations for the time distribution function of the quantum Bose system with a separate contribution from both the kinetic and potential energies of particle interactions are obtained. The generalized transport coefficients are determined accounting for the consistent description of kinetic and hydrodynamic processes.

  7. Controlled release systems containing solid dispersions: strategies and mechanisms.

    Science.gov (United States)

    Tran, Phuong Ha-Lien; Tran, Thao Truong-Dinh; Park, Jun Bom; Lee, Beom-Jin

    2011-10-01

    In addition to a number of highly soluble drugs, most new chemical entities under development are poorly water-soluble drugs generally characterized by an insufficient dissolution rate and a small absorption window, leading to the low bioavailability. Controlled-release (CR) formulations have several potential advantages over conventional dosage forms, such as providing a uniform and prolonged therapeutic effect to improve patient compliance, reducing the frequency of dosing, minimizing the number of side effects, and reducing the strength of the required dose while increasing the effectiveness of the drug. Solid dispersions (SD) can be used to enhance the dissolution rate of poorly water-soluble drugs and to sustain the drug release by choosing an appropriate carrier. Thus, a CR-SD comprises both functions of SD and CR for poorly water-soluble drugs. Such CR dosage forms containing SD provide an immediately available dose for an immediate action followed by a gradual and continuous release of subsequent doses to maintain the plasma concentration of poorly water-soluble drugs over an extended period of time. This review aims to summarize all currently known aspects of controlled release systems containing solid dispersions, focusing on the preparation methods, mechanisms of action and characterization of physicochemical properties of the system.

  8. Iodine removal in containment filtered venting system during nuclear accident

    International Nuclear Information System (INIS)

    Bera, Subrata; Deo, Anuj Kumar; Nagrale, D.B.; Paul, U.K.; Prasad, M.; Gaikwad, A.J.

    2015-01-01

    Post Fukushima nuclear accident, containment filtered venting system is being introduced in Indian nuclear power plant to strengthen the defense in depth safety barrier by depressurizing the containment building along with minimization of radioactivity release to environment during a severe accident. Radioactive iodine is one of the major contributors to radiation dose during early release phase of a severe accident. Physical and Chemical form of iodine and iodine bearing compounds includes particulates, elemental and organic. In the most efficient design of CFVS, wet scrubbing mechanism has been employed through use of venture scrubber. The Iodine removal process in wet scrubber involves two processes: chemical reaction in highly alkaline aqueous solution and impingement of particulates with water droplets produced in the venturi nozzle. In this paper, venturi has been modeled using the Calvert model. The variation of efficiency has been estimated for the different particle sizes. The impact of the shape parameter of log-normal distribution on the amount of scrubbed iodine has also been assessed. Release phase wise the scrubbed amount of iodine in the venturi based CFVS system has been estimated for a typical BWR. (author)

  9. AC-600 passive containment cooling system performance research

    International Nuclear Information System (INIS)

    Jia Baoshan; Yu Jiyang; Shi Junying

    1997-01-01

    a code named PCCSAC which is able to predict both the evaporating film on the outside surface of the vessel and the condensed film on its inside is developed successfully. It is a special software tool to analyze the passive containment cooling system (PCCS) performance in the design of AC-600. The author includes the establishment of physical models, selection of numerical methods, debugging and verification of the code and application of the code in the AC-600 PCCS. In physical models, the fundamental conservation equations about various areas and heat conduction equations are established. In order to make the equations to meet the closed form of solution, a lot of structure formulae are complemented. After repeated selection and demonstration of the numerical methods, the backward difference method Gear which is generally used for stiff problem is chosen for the solution of ordinary differential equations derived from the physical models. The results of standard example calculated by the PCCSAC code and the COMMIX code which is used to analyze westinghouse AP-600 are same in the main. The reliability and validity are verified from the calculations. The PCCSAC code is applied in the calculations of two important LOCA used in the containment safety analyses. The sensitivity of main parameters in the system based on LOCA are studied. All the results are reasonable and in agreement with the theoretical analyses. It can be concluded that the PCCSAC code is able to be used for the analyses of AC-600 PCCS performance

  10. MIDAS, prototype Multivariate Interactive Digital Analysis System, phase 1. Volume 1: System description

    Science.gov (United States)

    Kriegler, F. J.

    1974-01-01

    The MIDAS System is described as a third-generation fast multispectral recognition system able to keep pace with the large quantity and high rates of data acquisition from present and projected sensors. A principal objective of the MIDAS program is to provide a system well interfaced with the human operator and thus to obtain large overall reductions in turnaround time and significant gains in throughput. The hardware and software are described. The system contains a mini-computer to control the various high-speed processing elements in the data path, and a classifier which implements an all-digital prototype multivariate-Gaussian maximum likelihood decision algorithm operating at 200,000 pixels/sec. Sufficient hardware was developed to perform signature extraction from computer-compatible tapes, compute classifier coefficients, control the classifier operation, and diagnose operation.

  11. STUDIES ON VINYL POLYMERIZATION WITH INITIATION SYSTEM CONTAINING AMINE DERIVATIVES

    Institute of Scientific and Technical Information of China (English)

    QIU Kunyuan; ZHANG Jingyi; FENG Xinde(S. T. Voong)

    1983-01-01

    Two main types of amine-containing initiation systems were studied in this work. In the case of MMA polymerization initiated by BPO-amine (DMT, DHET, DMA) redox systems, it was found that the polymerization rate and colour stability of the polymer for different amine systems were in the following order: DMT≈DHET>DMA. Accordingly, BPO-DMT and BPO-DHET are effective initiators. In the case of MEMA polymerization by amine (DMT, DHET, DMA) alone, it was found that the polymerization rate and the percentage of conversion for these different amine systems were in the following order: DMT≥DHET>DMA. The polymerization rate and the percentage of conversion also increased with the increase of DMT concentration. From the kinetic investigation the rate equation of Rp=K [DMT]1/2 [MEMA]3/2 was obtained, and the overall activation energy of polymerization was calculated to be 34.3 KJ/mol (8.2 Kcal/mol). Moreover, the polymerization of MEMA in the presence of DMT was strongly inhibited by hydroquinone, indicating the polymerization being free radical in nature. From these results, the mechanism of MEMA polymerization initiated by amine was proposed.

  12. Computer systems and software description for Standard-E+ Hydrogen Monitoring System (SHMS-E+)

    International Nuclear Information System (INIS)

    Tate, D.D.

    1997-01-01

    The primary function of the Standard-E+ Hydrogen Monitoring System (SHMS-E+) is to determine tank vapor space gas composition and gas release rate, and to detect gas release events. Characterization of the gas composition is needed for safety analyses. The lower flammability limit, as well as the peak burn temperature and pressure, are dependent upon the gas composition. If there is little or no knowledge about the gas composition, safety analyses utilize compositions that yield the worst case in a deflagration or detonation. Knowledge of the true composition could lead to reductions in the assumptions and therefore there may be a potential for a reduction in controls and work restrictions. Also, knowledge of the actual composition will be required information for the analysis that is needed to remove tanks from the Watch List. Similarly, the rate of generation and release of gases is required information for performing safety analyses, developing controls, designing equipment, and closing safety issues. This report outlines the computer system design layout description for the Standard-E+ Hydrogen Monitoring System

  13. CONPAS 1.0 (CONtainment Performance Analysis System). User's manual

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Jin, Young Ho

    1996-04-01

    CONPAS (CONtainment Performance Analysis System) is a verified computer code package to integrate the numerical, graphical, and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically under a PC window environment. Compared with the existing DOS-based computer codes for Level 2 PSA, the most important merit of the window-based computer code is that user can easily describe and quantify the accident progression models, and manipulate the resultant outputs in a variety of ways. As a main logic for accident progression analysis, CONPAS employs a concept of the small containment phenomenological event tree (CPET) helpful to trace out visually individual accident progressions and of the large supporting event tree (LSET) for its detailed quantification. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules; (1) ET Editor for construction of several event tree models describing the accident progressions, (2) Computer for quantification of the constructed event trees and graphical display of the resultant outputs, (3) Text Editor for preparation of input decks for quanification and utilization of calculational results, and (4) Mechanistic Code Plotter for utilization of results obtained from severe accident analysis codes. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friend interface. 10 refs. (Author) .new

  14. FURNACE; a toroidal geometry neutronic program system method description and users manual

    International Nuclear Information System (INIS)

    Verschuur, K.A.

    1984-12-01

    The FURNACE program system performs neutronic and photonic calculations in 3D toroidal geometry for application to fusion reactors. The geometry description is quite general, allowing any torus cross section and any neutron source density distribution for the plasma, as well as simple parametric representations of circular, elliptic and D-shaped tori and plasmas. The numerical method is based on an approximate transport model that produces results with sufficient accuracy for reactor-design purposes, at acceptable calculational costs. A short description is given of the numerical method, and a user manual for the programs of the system: FURNACE, ANISN-PT, LIBRA, TAPEMA and DRAWER is presented

  15. Supramolecular effects in dendritic systems containing photoactive groups

    Directory of Open Access Journals (Sweden)

    GIANLUCA CAMILLO AZZELLINI

    2000-03-01

    Full Text Available In this article are described dendritic structures containing photoactive groups at the surface or in the core. The observed supramolecular effects can be attributed to the nature of the photoactive group and their location in the dendritic architecture. The peripheric azobenzene groups in these dendrimeric compounds can be regarded as single residues that retain the spectroscopic and photochemical properties of free azobenzene moiety. The E and Z forms of higher generation dendrimer, functionalized with azobenzene groups, show different host ability towards eosin dye, suggesting the possibility of using such dendrimer in photocontrolled host-guest systems. The photophysical properties of many dendritic-bipyridine ruthenium complexes have been investigated. Particularly in aerated medium more intense emission and a longer excited-state lifetime are observed as compared to the parent unsubstituted bipyridine ruthenium complexes. These differences can be attributed to a shielding effect towards dioxygen quenching originated by the dendritic branches.

  16. Design and hydrodynamic testing of an oil slick containment system

    International Nuclear Information System (INIS)

    Allen-Jones, J.

    1997-01-01

    Aspects of mechanical containment of spilled oil were studied. The focus was on design problems and the development of a model for global loading on a horizontal catenary of a previously defined form. The result is then compared with existing theoretical formulations and an approximate model is developed for the effect of flow through the system in deep water. The modified result is again compared with accepted formulations and with sea-trial data. The leading edge of the skirt was observed to oscillate sinusoidally. Experimental results obtained from pressure transducer data and calibrated underwater video measurements show that the oscillation period diminishes with increases in tow speed. In contrast, the magnitude of the oscillation increases while mean deviation from datum draught returns to zero. 14 refs., 5 tabs., 31 figs

  17. Thermodynamic Descriptions of NI Alloys Containing AL, CR, and RU: A Computational Thermodynamic Approach Coupled with Experiments

    National Research Council Canada - National Science Library

    Chang, Y. A

    2006-01-01

    .... In addition, we extended our effort to include the use of the Cluster/Site Approximation (CSA) to describe the fcc phases in the disordered and ordered states such as the prototype Cu-Ag-Au in 2003/2004 and then I in real ternary systems in 2004/2005...

  18. A description of the apparatus to be used in interaction experiments with the ABC laser system

    International Nuclear Information System (INIS)

    Caruso, A.; Strangio, M.; Andreoli, P.L.; Cerioni, I.; Di Paolo, A.; Di Virgilio, L.

    1988-01-01

    This report contains the part of the Frascati Laboratorio Fusione Laser activity related to the Apparatus (target chamber, position and alignement system, diagnostics) to be used in the interaction experiments with the ABC laser system

  19. On the description of classical Einstein relativistic two-particle systems

    International Nuclear Information System (INIS)

    Aaberge, T.

    1978-01-01

    The author starts by considering the system of one free particle, and gives a sufficiently general description of this system to include the center of mass of systems of several particles. He then passes to the system of two particles. The coordinates separating the center of mass and the internal system are defined and the dynamics discussed. Finally the author outlines the construction of a more restrictive two-particle theory, and studies some consequences of the definition of a particle in an external field as a two-particle system in the limit where the mass of one of the particles becomes infinite. (Auth.)

  20. Characterization and stability studies of emulsion systems containing pumice

    Directory of Open Access Journals (Sweden)

    Marilene Estanqueiro

    2014-04-01

    Full Text Available Emulsions are the most common form of skin care products. However, these systems may exhibit some instability. Therefore, when developing emulsions for topical application it is interesting to verify whether they have suitable physical and mechanical characteristics and further assess their stability. The aim of this work was to study the stability of emulsion systems, which varied in the proportion of the emulsifying agent cetearyl alcohol (and sodium lauryl sulfate (and sodium cetearyl sulfate (LSX, the nature of the oily phase (decyl oleate, cyclomethicone or dimethicone and the presence or absence of pumice (5% w/w. While maintaining the samples at room temperature, rheology studies, texture analysis and microscopic observation of formulations with and without pumice were performed. Samples were also submitted to an accelerated stability study by centrifugation and to a thermal stress test. Through the testing, it was found that the amount of emulsifying agent affects the consistency and textural properties such as firmness and adhesiveness. So, formulations containing LSX (5% w/w and decyl oleate or dimethicone as oily phase had a better consistency and remained stable with time, so exhibited the best features to be used for skin care products.

  1. TRAC analysis of passive containment cooling system performance

    International Nuclear Information System (INIS)

    Arai, Kenji; Kataoka, Kazuyoshi; Nagasaka, Hideo

    1993-01-01

    A passive containment cooling system (PCCS) is a promising concept to improve the reliability of decay heat removal during an accident. Toshiba has carried out analytical studies for PCCS development in addition to experimental studies, using a best estimate thermal hydraulic computer code TRAC. In order to establish an analytical model for the PCCS performance analysis, it is necessary for the analytical model to be qualified against experimental results and thoroughly address the phenomena important for PCCS performance analysis. In this paper, the TRAC qualification for PCCS application is reported. A TRAC model has been verified against a drain line break simulation test conducted at the PCCS integral test facility, GIRAFFE. The result shows that the TRAC model can accurately predict the major system response and the PCCS performance in the drain line break test. In addition, the results of several sensitivity analyses, showing various points concerning the modeling in the PCCS performance analysis, have been reported. The analyses have been carried out for the SBWR and the analytical points are closely related to important phenomena which can affect PCCS performance

  2. Humos monitoring system of leaks in to the containment atmosphere

    International Nuclear Information System (INIS)

    Matal, O.; Zaloudek, J.; Matal, O. Jr.; Klinga, J.; Brom, J.

    1997-01-01

    HUmidity MOnitoring System (HUMOS) has been developed and designed to detect the presence of leak in selected primary circuit high energy pipelines and components that are evaluated from the point of view of Leak Before Break (LBB) requirements. It also requires to apply technical tools for detection and identification of coolant leaks from primary circuit and components of PWRs reactors. Safety significant of leaks depend on: leak source (location); leak rate, and leak duration. Therefore to detect and monitor coolant leaks in to the containment atmosphere during reactor operation is one of important safety measures. As potential leak sources flange connection in the upper head region of WWER reactors can be considered. HUMOS does not rely on the release of radioactivity to detect leaks but rather the relies on detection of moisture in the air resulting from a primary boundary leak. Because HUMOS relies on moisture and temperature detection, leaks can be detected without requiring the reactor to be critical. Therefore leaks can be detected during integrity pressure tests and any other mode of operation provided the reactor ventilation system is operating and primary circuit and components are pressurized. 3 figs

  3. A comparison of hardware description languages. [describing digital systems structure and behavior to a computer

    Science.gov (United States)

    Shiva, S. G.

    1978-01-01

    Several high level languages which evolved over the past few years for describing and simulating the structure and behavior of digital systems, on digital computers are assessed. The characteristics of the four prominent languages (CDL, DDL, AHPL, ISP) are summarized. A criterion for selecting a suitable hardware description language for use in an automatic integrated circuit design environment is provided.

  4. New Approach for Description of Sorption and Swelling phenomena in Liquid + Polymer Membrane Systems.

    Czech Academy of Sciences Publication Activity Database

    Randová, A.; Bartovská, L.; Hovorka, Š.; Bartovský, T.; Izák, Pavel; Kárászová, Magda; Vopička, O.; Lindnerová, V.

    2017-01-01

    Roč. 179, MAY (2017), s. 475-485 ISSN 1383-5866 R&D Projects: GA MŠk(CZ) LD14094 Institutional support: RVO:67985858 Keywords : description of sorption * polymer membranes systems * new method Subject RIV: CI - Industrial Chemistry, Chemical Engineering OBOR OECD: Chemical process engineering Impact factor: 3.359, year: 2016

  5. The nonlinear finite element analysis program NUCAS (NUclear Containment Analysis System) for reinforced concrete containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Jin; Lee, Hong Pyo; Seo, Jeong Moon [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    The maim goal of this research is to develop a nonlinear finite element analysis program NUCAS to accurately predict global and local failure modes of containment building subjected to internal pressure. In this report, we describe the techniques we developed throught this research. An adequate model to the analysis of containment building such as microscopic material model is adopted and it applied into the development Reissner-Mindlin degenerated shell element. To avoid finite element deficiencies, the substitute strains based on the assumed strain method is used in the shell formulation. Arc-length control method is also adopted to fully trace the peak load-displacement path due to crack formation. In addition, a benchmark test suite is developed to investigate the performance of NUCAS and proposed as the future benchmark tests for nonlinear analysis of reinforced concrete. Finally, the input format of NUCAS and the examples of input/output file are described. 39 refs., 65 figs., 8 tabs. (Author)

  6. Simulation of the containment spray system test PACOS PX2.2 with the integral code ASTEC and the containment code system COCOSYS

    International Nuclear Information System (INIS)

    Risken, Tobias; Koch, Marco K.

    2011-01-01

    The reactor safety research contains the analysis of postulated accidents in nuclear power plants (npp). These accidents may involve a loss of coolant from the nuclear plant's reactor coolant system, during which heat and pressure within the containment are increased. To handle these atmospheric conditions, containment spray systems are installed in various light water reactors (LWR) worldwide as a part of the accident management system. For the improvement and the safety ensurance in npp operation and accident management, numeric simulations of postulated accident scenarios are performed. The presented calculations regard the predictability of the containment spray system's effect with the integral code ASTEC and the containment code system COCOSYS, performed at Ruhr-Universitaet Bochum. Therefore the test PACOS Px2.2 is simulated, in which water is sprayed in the stratified containment atmosphere of the BMC (Battelle Modell-Containment). (orig.)

  7. Description of leakage monitoring system at Angra 2 nuclear power plant primary circuit

    International Nuclear Information System (INIS)

    Costa, Lilian Rose Sobral da; Mendes, Jorge Eduardo de Souza

    1999-01-01

    This paper describes the Leakage Monitoring System installed in Angra 2 NPP. This system has the task of detecting, localizing and quantifying leaks in systems for which rupture preclusion is cited. These systems include the reactor coolant pressure boundary, the main steam and feedwater lines within the containment, and the main steam safety and relief valve station in the valve annex. (author)

  8. Using Self-Description to Handle Change in Systems, CMS-CR-2002-010

    CERN Document Server

    Estrella, F; Le Goff, J M; McClatchey, R; Murray, S

    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...

  9. Observing Participating Observation—A Re-description Based on Systems Theory

    Directory of Open Access Journals (Sweden)

    Tina Bering Keiding

    2010-09-01

    Full Text Available Current methodology concerning participating observation in general leaves the act of observation unobserved. Approaching participating observation from systems theory offers fundamental new insights into the topic. Observation is always participation. There is no way to escape becoming a participant and, as such, co-producer of the observed phenomenon. There is no such thing as a neutral or objective description. As observation deals with differences and process meaning, all descriptions are re-constructions and interpretations of the observed. Hence, the idea of neutral descriptions as well as the idea of the naïve observer becomes a void. Not recognizing and observing oneself as observer and co-producer of empirical data simply leaves the process of observation as the major unobserved absorber of contingency in data production based on participating observation. URN: urn:nbn:de:0114-fqs1003119

  10. A Fundamental Scale of Descriptions for Analyzing Information Content of Communication Systems

    Directory of Open Access Journals (Sweden)

    Gerardo Febres

    2015-03-01

    Full Text Available The complexity of the description of a system is a function of the entropy of its symbolic description. Prior to computing the entropy of the system’s description, an observation scale has to be assumed. In texts written in artificial and natural languages, typical scales are binary, characters, and words. However, considering languages as structures built around certain preconceived set of symbols, like words or characters, limits the level of complexity that can be revealed analytically. This study introduces the notion of the fundamental description scale to analyze the essence of the structure of a language. The concept of Fundamental Scale is tested for English and musical instrument digital interface (MIDI music texts using an algorithm developed to split a text in a collection of sets of symbols that minimizes the observed entropy of the system. This Fundamental Scale reflects more details of the complexity of the language than using bits, characters or words. Results show that this Fundamental Scale allows to compare completely different languages, such as English and MIDI coded music regarding its structural entropy. This comparative power facilitates the study of the complexity of the structure of different communication systems.

  11. Systems Biology Graphical Notation: Process Description language Level 1 Version 1.3.

    Science.gov (United States)

    Moodie, Stuart; Le Novère, Nicolas; Demir, Emek; Mi, Huaiyu; Villéger, Alice

    2015-09-04

    The Systems Biological Graphical Notation (SBGN) is an international community effort for standardized graphical representations of biological pathways and networks. The goal of SBGN is to provide unambiguous pathway and network maps for readers with different scientific backgrounds as well as to support efficient and accurate exchange of biological knowledge between different research communities, industry, and other players in systems biology. Three SBGN languages, Process Description (PD), Entity Relationship (ER) and Activity Flow (AF), allow for the representation of different aspects of biological and biochemical systems at different levels of detail. The SBGN Process Description language represents biological entities and processes between these entities within a network. SBGN PD focuses on the mechanistic description and temporal dependencies of biological interactions and transformations. The nodes (elements) are split into entity nodes describing, e.g., metabolites, proteins, genes and complexes, and process nodes describing, e.g., reactions and associations. The edges (connections) provide descriptions of relationships (or influences) between the nodes, such as consumption, production, stimulation and inhibition. Among all three languages of SBGN, PD is the closest to metabolic and regulatory pathways in biological literature and textbooks, but its well-defined semantics offer a superior precision in expressing biological knowledge.

  12. Smart integrated containment leakage rate test system using wireless communication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Hwan; Lee, Sang Yong; Kim, Jung Sun; Kim, Gun Soo; Kim, Jong Myeong; Ahn, Jong Han [Research and Development Center, Ulsan (Korea, Republic of)

    2012-10-15

    Integrated Leakage Rate Test (ILRT) is the important test the confidentiality and integrity of the containment building, which is the last barrier when Design basis accidents (DBA) of Nuclear Power plant occur. Since the result of this test is the basis to guarantee the safety of nuclear power plants, the test process, test procedure, and the test equipment are required to have high reliability. The test devices previously used have been products of VOLUMERTRICS and GRAFTEL of USA. These devices have been inconvenient to calibrate and use. Thus improved devices needed to be developed to remove the inconveniences, to verify the safety of Korean nuclear power plants with Korea's own technology, and to secure core technology. A new leak test system was developed by domestic technology for that purpose and needed to be verified. In this paper, technical details of the newly developed easy to use and highly reliable measuring test device, which is in operation at the nuclear power plant sites, will be introduced. State of art technology was applied to the device to address the shortcomings of previous US made devices and the difficulties to use on site.

  13. Fuel salt and container material studies for MOSART transforming system

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A. [National Research Center, Kurchatov Institute, Moscow (Russian Federation); Afonichkin, V.; Bovet, A.; Khokhlov, V. [Institute of High Temperature Electrochemisty, Ekaterinburg (Russian Federation); Subbotin, V.; Gordeev, M.; Panov, A.; Toropov, A. [Institute of Technical Physics, Snezhinsk (Russian Federation)

    2013-07-01

    A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

  14. Throughput Evaluation of an Autonomous Sustainment Cargo Container System

    National Research Council Canada - National Science Library

    Yeh, Mingtze

    2007-01-01

    .... Autonomous containers will play an essential role in the ability to deliver logistical supplies to waterborne littoral vessels enabling them to maintain station and complete there military operations...

  15. Design of database management system for 60Co container inspection system

    International Nuclear Information System (INIS)

    Liu Jinhui; Wu Zhifang

    2007-01-01

    The function of the database management system has been designed according to the features of cobalt-60 container inspection system. And the software related to the function has been constructed. The database querying and searching are included in the software. The database operation program is constructed based on Microsoft SQL server and Visual C ++ under Windows 2000. The software realizes database querying, image and graph displaying, statistic, report form and its printing, interface designing, etc. The software is powerful and flexible for operation and information querying. And it has been successfully used in the real database management system of cobalt-60 container inspection system. (authors)

  16. Description langugage for the modelling and analysis of temporal change of instrumentation and control system structures

    International Nuclear Information System (INIS)

    Goering, Markus Heinrich

    2013-01-01

    comprehensive requirements for the superposition of event sequences and failure combinations. For this reason, the synthesis of a description language, under consideration of the aforementioned challenges, is necessary; supplemented by a method utilising the description language for efficient engineering and I and C design analysis. Due to the abstraction of AutomationML as a meta-metamodel, it is utilised as the basis for the description language synthesis, however AutomationML also does not fulfil all target criteria. On one hand, the description language synthesis is based on the syntax of AutomationML, and on the other hand, the semantics are determined by the context of computer-based I and C in nuclear power plants and structured utilising the general product, function, and location structures of standard IEC 81346. When modelling I and C failure combinations in addition to event sequences, the description language is completed by an event structure, for which CDL is utilised for the conceptualisation and the formalisation is accomplished with PSL. The PSL formalisation allows for implementing the description language in a knowledge-based system, so that automated engineering is enabled. The I and C modelling, as part of the method, is embedded in the IEC 61513 I and C safety life-cycle and is realised in two steps. Consequently, the I and C design can be analysed at both plant and system level. For the I and C design analysis the concepts of the D3-analysis, FTA, ETA, and FMEA are combined. The thesis is concluded with an example applying the description language and method to the modernisation of a reactor protection system; this illustrates the validation of the overall concept developed in this thesis.

  17. Description langugage for the modelling and analysis of temporal change of instrumentation and control system structures

    Energy Technology Data Exchange (ETDEWEB)

    Goering, Markus Heinrich

    2013-10-25

    comprehensive requirements for the superposition of event sequences and failure combinations. For this reason, the synthesis of a description language, under consideration of the aforementioned challenges, is necessary; supplemented by a method utilising the description language for efficient engineering and I and C design analysis. Due to the abstraction of AutomationML as a meta-metamodel, it is utilised as the basis for the description language synthesis, however AutomationML also does not fulfil all target criteria. On one hand, the description language synthesis is based on the syntax of AutomationML, and on the other hand, the semantics are determined by the context of computer-based I and C in nuclear power plants and structured utilising the general product, function, and location structures of standard IEC 81346. When modelling I and C failure combinations in addition to event sequences, the description language is completed by an event structure, for which CDL is utilised for the conceptualisation and the formalisation is accomplished with PSL. The PSL formalisation allows for implementing the description language in a knowledge-based system, so that automated engineering is enabled. The I and C modelling, as part of the method, is embedded in the IEC 61513 I and C safety life-cycle and is realised in two steps. Consequently, the I and C design can be analysed at both plant and system level. For the I and C design analysis the concepts of the D3-analysis, FTA, ETA, and FMEA are combined. The thesis is concluded with an example applying the description language and method to the modernisation of a reactor protection system; this illustrates the validation of the overall concept developed in this thesis.

  18. Project W-211 Initial Tank Retrieval Systems (ITRS) Description of Operations for 241-AZ-102

    Energy Technology Data Exchange (ETDEWEB)

    BRIGGS, S.R.

    2000-02-25

    The primary purpose of the Initial Tank Retrieval Systems (ITRS) is to provide systems for retrieval of radioactive wastes stored in underground double-shell tanks (DSTs) for transfer to alternate storage, evaporation, pretreatment or treatment, while concurrently reducing risks associated with safety watch list and other DSTs. This Description of Operation (DOO) defines the control philosophy for the waste retrieval system for Tank 241-AZ-102 (AZ-102). This DOO provides a basis for the detailed design of the Project W-211 Retrieval Control System (RCS) for AZ-102 and also establishes test criteria for the RCS.

  19. Project W-211 Initial Tank Retrieval Systems (ITRS) Description of Operations for 241-AZ-102

    International Nuclear Information System (INIS)

    BRIGGS, S.R.

    2000-01-01

    The primary purpose of the Initial Tank Retrieval Systems (ITRS) is to provide systems for retrieval of radioactive wastes stored in underground double-shell tanks (DSTs) for transfer to alternate storage, evaporation, pretreatment or treatment, while concurrently reducing risks associated with safety watch list and other DSTs. This Description of Operation (DOO) defines the control philosophy for the waste retrieval system for Tank 241-AZ-102 (AZ-102). This DOO provides a basis for the detailed design of the Project W-211 Retrieval Control System (RCS) for AZ-102 and also establishes test criteria for the RCS

  20. Qualitative Description and Quantitative Optimization of Tactical Reconnaissance Agents System Organization

    Directory of Open Access Journals (Sweden)

    Xiong Li

    2012-08-01

    Full Text Available In this paper, the problem of qualitative description and quantitative optimization for tactical reconnaissance agents system organization is considered with objective of higher teamwork efficiency and more reasonable task balancing strategies. By analyzing tactical reconnaissance system and its environment, task-(role-entity agent mapping mechanism and agents in system organization, the system framework is qualitatively described. By transforming the system into an interaction task request-service mechanism queuing system, a Markov chain of system state transition is obtained, since its state transition process in interaction is Markov process and accords with real tactical reconnaissance behaviors. By solving the state transition equations, the inherent relationship of tactical reconnaissance agents is found and the optimized system configuration is obtained. The established simulation demonstration system proves that the proposed approach and model are feasible and effective.

  1. The water reactor safety research project index: a description of the computerized system for its databank

    International Nuclear Information System (INIS)

    Della Loggia, E.; Primavera, R.

    1993-01-01

    The water reactor nuclear safety research project index has been published by the CEC for many years as a compilation of information on research projects relating to LWR nuclear safety. Since 1981, it has been published, alternatively with NEA (OECD), every second year. The number of contributions from research organizations in Community member countries has steadily increased and reached the level of 1 700 pages, in which more than 600 project descriptions have been collected. In 1988, for the first time, the document was produced using a computerized system developed with the assistance of the ISEI (Institute for Systems Engineering and Informatics) of JRC Ispra. The data have been stored in a computer based in Ispra. The system allows searching a preselected set of subjects through the information stored in the computer: it makes the updating of the projects description much easier and makes the retrieval of the data possible. This report presents a short description of the computerized system developed for the databank of the index. The computerized system presented in this report is structured in a quite general way and for that reason can be adapted very easily to every field where a databank needs to be constituted in order to collect extended information on several projects. (authors). 6 figs., 5 tabs., 5 refs

  2. Probabilistic endowment appraisal system based upon the formalization of geologic decisions. General description

    International Nuclear Information System (INIS)

    Harris, D.P.; Carrigan, F.J.

    1980-04-01

    The objectives of this study include the design of an appraisal system which has the following features: estimates uranium endowment, not resources; formalizes the geologist's geoscience and assists the geologist in the exercise of his geoscience; describes the probability distribution for uranium endowment; diminishes or at least does not contribute to psychometric biases; provides for anonymous exchange among multiple experts of tenets of geoscience, but not the exchange of endowment estimates; provides an endowment estimate based upon geoscience only; is not easily gamed or manipulated; and provides for a quick and easy review of geoscience and resource information. This report is reflective of its title, a general description. The appraisal system resulting from this research is complex in the detail of its design and use. However, the major concepts which are reflected by the system are simple. The purpose of this report is to establish clearly these major concepts and the manner in which the system applies these concepts. Many details, refinements, and caveats are purposefully suppressed in order to provide this general description. While this suppression is a loss to some readers, it is a benefit to a wider spectrum of readers. Those interested in the nuts and bolts of the system will also want to read the user's manual which accompanies this general description

  3. Verifying Real-Time Systems using Explicit-time Description Methods

    Directory of Open Access Journals (Sweden)

    Hao Wang

    2009-12-01

    Full Text Available Timed model checking has been extensively researched in recent years. Many new formalisms with time extensions and tools based on them have been presented. On the other hand, Explicit-Time Description Methods aim to verify real-time systems with general untimed model checkers. Lamport presented an explicit-time description method using a clock-ticking process (Tick to simulate the passage of time together with a group of global variables for time requirements. This paper proposes a new explicit-time description method with no reliance on global variables. Instead, it uses rendezvous synchronization steps between the Tick process and each system process to simulate time. This new method achieves better modularity and facilitates usage of more complex timing constraints. The two explicit-time description methods are implemented in DIVINE, a well-known distributed-memory model checker. Preliminary experiment results show that our new method, with better modularity, is comparable to Lamport's method with respect to time and memory efficiency.

  4. Microparticulate drug delivery system containing tramadol hydrochloride for pain treatment.

    Science.gov (United States)

    Ciurba, Adriana; Todoran, Nicoleta; Vari, C E; Lazăr, Luminita; Al Hussein, Stela; Hancu, G

    2014-01-01

    The current trend of replacing conventional pharmaceutical forms is justified because most substances administered in this form give fluctuations of therapeutic concentrations and often outside the therapeutic range. In addition, these formulations offer a reduction in the dose or the number of administrations, thus increasing patient compliance. In the experiment, we developed an appropriate technology for the preparation of gelatin microspheres containing tramadol hydrochloride by emulsification/cross-linking method. The formulated microspheres were characterized by product yield, size distribution, encapsulation efficiency and in vitro release of tramadol hydrochloride. Data obtained from in vitro release studies were fitted to various mathematical models to elucidate the transport mechanisms. The kinetic models used were zero-order, first-order, Higuchi Korsmeyer-Peppas and Hopfenberg. Spherical microspheres were obtained, with free-flowing properties. The entrapment efficiency of tramadol hydrochloride in microparticles was 79.91% and product yield -94.92%. As the microsphere size was increased, the entrapment efficiency increased. This was 67.56, 70.03, 79.91% for formulations MT80-250, MT8-500 and, MT250-500. High entrapment efficiency was observed for MT250-500 formulation. The gelatin microspheres had particle sizes ranging from 80 to 500 microm. The drug was released for a period of 12 hours with a maximum release of 96.02%. Of the three proposed formulations, MT250-500 presented desirable properties and optimal characteristics for the therapy of pain. Release of tramadol hydrochloridi was best fitted to Korsmeyer-Peppas equation because the Akaike Information Criterion had the lowest values for this kinetic model. These results suggest the opportunity to influence the therapeutic characteristics of gelatin microspheres to obtain a suitable drug delivery system for the oral administration of tramadol hydrochloride.

  5. The status of the Bubbler Condenser Containment System for the Reactors of the VVER-440/213 Type

    International Nuclear Information System (INIS)

    Karwat, H.; Rosinger, H.E.

    1998-01-01

    VVER-440/213 Pressurized Water Reactors have a pressure-suppression containment structure called a 'Bubbler Condenser' tower which can reduce the design pressure of the entire containment following a design basis accident (DBA), such as a loss-of-coolant accident (LOCA). The bubbler condenser pressure suppression system provides reduction of the LOCA containment pressure by the condensation of released steam in a water pool. World-wide there are 14 nuclear power plants of the VVER-440/213 type in Eastern Europe and Russia. One of the safety concerns for the VVER-440/213 reactors relates to the ability of the bubbler condenser containment system to function satisfactorily and to maintain its integrity following certain postulated accidents and thus limit the release of radioactive material to the environment. The complicated geometry of the bubbler condenser unit, and the dependence on several moving devices and interlocks are the main doubts expressed by different specialists with regard to the design. General description of the bubbler condenser containment system, the physical processes, concerns and design assessment of the bubbler condenser containment system, presentation of the OECD's Unified Bubbler Condenser Research Project (UBCRP) and the European Commission PHARE/TACIS project. Recent utility investigations are also discussed

  6. Exploratory Shaft, Phase 1, Project B-314: Title 1 design report system design description

    International Nuclear Information System (INIS)

    Hanlen, D.F.

    1983-01-01

    The report describes the project and the project systems, the principal design bases, and principal hazards and project interfaces. This report also contains the Title 1 Estimate Summary. 5 figs., 8 tabs

  7. Image processing in 60Co container inspection system

    International Nuclear Information System (INIS)

    Wu Zhifang; Zhou Liye; Wang Liqiang; Liu Ximing

    1999-01-01

    The authors analyzes the features of 60 Co container inspection image, the design of several special processing methods for container image and some normal processing methods for two-dimensional digital image, including gray enhancement, pseudo-enhancement, space filter, edge enhancement, geometry process, etc. It gives out the way to carry out the above mentioned process in Windows 95 or Win NT. It discusses some ways to improve the image processing speed on microcomputer and good results were obtained

  8. CIRCUIT IMPLEMENTATION OF VHDL-DESCRIPTIONS OF SYSTEMS OF PARTIAL BOOLEAN FUNCTIONS

    Directory of Open Access Journals (Sweden)

    P. N. Bibilo

    2016-01-01

    Full Text Available Method for description of incompletely specified (partial Boolean functions in VHDL is proposed. Examples of synthesized VHDL models of partial Boolean functions are presented; and the results of experiments on circuit implementation of VHDL descriptions of systems of partial functions. The realizability of original partial functions in logical circuits was verified by formal verification. The results of the experiments show that the preliminary minimization in DNF class and in the class of BDD representations for pseudo-random systems of completely specified functions does not improve practically (and in the case of BDD sometimes worsens the results of the subsequent synthesis in the basis of FPGA unlike the significant efficiency of these procedures for the synthesis of benchmark circuits taken from the practice of the design.

  9. System requirements and design description for the environmental requirements management interface (ERMI)

    International Nuclear Information System (INIS)

    Biebesheimer, E.

    1997-01-01

    This document describes system requirements and the design description for the Environmental Requirements Management Interface (ERMI). The ERMI database assists Tank Farm personnel with scheduling, planning, and documenting procedure compliance, performance verification, and selected corrective action tracking activities for Tank Farm S/RID requirements. The ERMI database was developed by Science Applications International Corporation (SAIC). This document was prepared by SAIC and edited by LMHC

  10. Uniform description and access to Knowledge Organization Systems with BARTOC and JSKOS

    OpenAIRE

    Voß, Jakob; Ledl, Andreas; Balakrishnan, Uma

    2016-01-01

    The Basel Register of Thesauri, Ontologies & Classifications (BARTOC) provides information about a large number of Knowledge Organization Systems (KOS) such as classifications, thesauri, authority files etc. To further improve availability and usefulness of both the description and content of KOS, they are mapped to the uniform JSKOS data format being developed in project coli-conc. Specification of a corresponding JSKOS-API will allow users to directly browse and search in KOS from any place.

  11. Evaluation of an air drilling cuttings containment system

    Energy Technology Data Exchange (ETDEWEB)

    Westmoreland, J.

    1994-04-01

    Drilling at hazardous waste sites for environmental remediation or monitoring requires containment of all drilling fluids and cuttings to protect personnel and the environment. At many sites, air drilling techniques have advantages over other drilling methods, requiring effective filtering and containment of the return air/cuttings stream. A study of. current containment methods indicated improvements could be made in the filtering of radionuclides and volatile organic compounds, and in equipment like alarms, instrumentation or pressure safety features. Sandia National Laboratories, Dept. 61 11 Environmental Drilling Projects Group, initiated this work to address these concerns. A look at the industry showed that asbestos abatement equipment could be adapted for containment and filtration of air drilling returns. An industry manufacturer was selected to build a prototype machine. The machine was leased and put through a six-month testing and evaluation period at Sandia National Laboratories. Various materials were vacuumed and filtered with the machine during this time. In addition, it was used in an actual air drive drilling operation. Results of these tests indicate that the vacuum/filter unit will meet or exceed our drilling requirements. This vacuum/filter unit could be employed at a hazardous waste site or any site where drilling operations require cuttings and air containment.

  12. 40 CFR 281.37 - Financial responsibility for UST systems containing petroleum.

    Science.gov (United States)

    2010-07-01

    ... systems containing petroleum. 281.37 Section 281.37 Protection of Environment ENVIRONMENTAL PROTECTION... for No-Less-Stringent § 281.37 Financial responsibility for UST systems containing petroleum. (a) In... UST systems containing petroleum, the state requirements for financial responsibility for petroleum...

  13. Materials performance in off-gas systems containing iodine

    International Nuclear Information System (INIS)

    Beavers, J.A.; Berry, W.E.; Griess, J.C.

    1981-11-01

    During the reprocessing of spent reactor fuel elements, iodine is released to gas streams from which it is ultimately removed by conversion to nonvolatile iodic acid. Under some conditions iodine can produce severe corrosion in off-gas lines; in this study these conditions were established. Iron- and nickel-based alloys containing more than 6% molybdenum, such as Hastelloy G (7%), Inconel 625 (9%), and Hastelloy C-276 (16%), as well as titanium and zirconium, remained free of attack under all conditions tested. When the other materials, notably the austenitic stainless steels, were exposed to gas streams containing even only low concentrations of iodine and water vapors at 25 and 40 0 C, a highly corrosive, brownish-green liquid formed on their surfaces. In the complete absence of water vapor, the iodine-containing liquid did not form and all materials remained unaffected. The liquid that formed had a low pH (usually 2 inhibited attack

  14. The descriptive properties of prescriptive theories: an application of systems thinking in data warehousing

    Directory of Open Access Journals (Sweden)

    Roelien Goede

    2012-12-01

    Full Text Available Information systems and in particular data warehouses are very expensive systems to develop. It is therefore not advisable to experiment with ideas too different from current practices. This makes it difficult to apply prescriptive theories in an existing field. From theoretical considerations one might want to develop a data warehouse according to another method such as critical systems thinking methodology. It is however very difficult to persuade data warehouse practitioners to attempt such an experiment. This might be because they would rather adhere to known practices or that they are not sufficiently knowledgeable on critical systems thinking (or any other prescriptive theory to apply it to such an expensive project. This paper describes a method in which prescriptive theories may be used descriptively to analyse their applicability in a specific field of application. The proposed method is used to understand the practices of the data warehouse discipline from the perspectives of the systems thinking discipline. It is also indicated how this method could be used in other studies where the behaviour of participants is viewed from a point of view of which the detail are unknown to the participants. Keywords: Data warehousing, Systems thinking, Prescriptive theory, Descriptive theory, Interpretative research. Disciplines: Information technology, systems theory, data warehousing, hermeneutics

  15. Analytical studies related to Indian PHWR containment system performance

    International Nuclear Information System (INIS)

    Haware, S.K.; Markandeya, S.G.; Ghosh, A.K.; Kushwaha, H.S.; Venkat Raj, V.

    1998-01-01

    Build-up of pressure in a multi-compartment containment after a postulated accident, the growth, transportation and removal of aerosols in the containment are complex processes of vital importance in deciding the source term. The release of hydrogen and its combustion increases the overpressure. In order to analyze these complex processes and to enable proper estimation of the source term, well tested analytical tools are necessary. This paper gives a detailed account of the analytical tools developed/adapted for PSA level 2 studies. (author)

  16. Interim reliability-evaluation program: analysis of the Browns Ferry, Unit 1, nuclear plant. Appendix B - system descriptions and fault trees

    International Nuclear Information System (INIS)

    Mays, S.E.; Poloski, J.P.; Sullivan, W.H.; Trainer, J.E.; Bertucio, R.C.; Leahy, T.J.

    1982-07-01

    This report describes a risk study of the Browns Ferry, Unit 1, nuclear plant. The study is one of four such studies sponsored by the NRC Office of Research, Division of Risk Assessment, as part of its Interim Reliability Evaluation Program (IREP), Phase II. This report is contained in four volumes: a main report and three appendixes. Appendix B provides a description of Browns Ferry, Unit 1, plant systems and the failure evaluation of those systems as they apply to accidents at Browns Ferry. Information is presented concerning front-line system fault analysis; support system fault analysis; human error models and probabilities; and generic control circuit analyses

  17. Lipid containing nanodrug delivery system for the treatment of Tuberculosis

    CSIR Research Space (South Africa)

    Lemmer, Yolandy

    2010-09-01

    Full Text Available of the antibiotics in the cells, hence reducing the dose frequency and simultaneously improve patient compliance. The cell wall envelope of Mycobacterium tuberculosis (M.tb) contains unique high molecular weight lipids. Of these, the most abundant are mycolic acids...

  18. Description of the map board portion of the Security Operations Center of the Plutonium Protection System

    International Nuclear Information System (INIS)

    Ringler, C.E.

    1979-05-01

    This report describes the console map board which is part of the Sandia-designed Plutonium Protection System tested at the Hanford Works. The board displays areas under surveillance and contains alarm lights and switches for communicating with the system's computer

  19. Collective and boson mapping description of a system of N Josephson junctions in a resonant cavity

    International Nuclear Information System (INIS)

    Ballesteros, A.; Civitarese, O.; Herranz, F.J.; Reboiro, M.

    2003-01-01

    A system of N two-level Josephson junctions, interacting between themselves and with a single-mode cavity field, is described in terms of the superposition of fermionic and bosonic excitations. The results of the exact diagonalization are compared with the results of the Tamm-Dancoff approximation and with the results of a boson mapping. It is found that the boson mapping provides a suitable description of the spectrum, sum rules, and response function of the system. The dependence of the results upon the number of junctions, the excitation of the cavity modes, and the coupling strengths is investigated

  20. SADE: system of acquisition of experimental data. Definition and analysis of an experiment description language

    International Nuclear Information System (INIS)

    Gagniere, Jean-Michel

    1983-01-01

    This research thesis presents a computer system for the acquisition of experimental data. It is aimed at acquiring, at processing and at storing information from particle detectors. The acquisition configuration is described by an experiment description language. The system comprises a lexical analyser, a syntactic analyser, a translator, and a data processing module. It also comprises a control language and a statistics management and plotting module. The translator builds up series of tables which allow, during an experiment, different sequences to be executed: experiment running, calculations to be performed on this data, building up of statistics. Short execution time and ease of use are always looked for [fr

  1. Knowledge-based driver assistance systems traffic situation description and situation feature relevance

    CERN Document Server

    Huelsen, Michael

    2014-01-01

    The comprehension of a traffic situation plays a major role in driving a vehicle. Interpretable information forms a basis for future projection, decision making and action performing, such as navigating, maneuvering and driving control. Michael Huelsen provides an ontology-based generic traffic situation description capable of supplying various advanced driver assistance systems with relevant information about the current traffic situation of a vehicle and its environment. These systems are enabled to perform reasonable actions and approach visionary goals such as injury and accident free driv

  2. [Systemic inflammation: theoretical and methodological approaches to description of general pathological process model. Part 3. Backgroung for nonsyndromic approach].

    Science.gov (United States)

    Gusev, E Yu; Chereshnev, V A

    2013-01-01

    Theoretical and methodological approaches to description of systemic inflammation as general pathological process are discussed. It is shown, that there is a need of integration of wide range of types of researches to develop a model of systemic inflammation.

  3. Summary description of the Babcock and Wilcox integrated nuclear design system

    International Nuclear Information System (INIS)

    Wittkopf, W.A.

    1976-03-01

    The Babcock and Wilcox integrated nuclear design system is divided into three broad areas: basic nuclear data processing, applications data processing, and nuclear design calculations. In basic nuclear data processing, basic nuclear data are collected, evaluated, and processed into a specified fine-energy mesh multigroup data file called a Master Library. In applications data processing, data for selected materials are retrieved from the Master Library and processed into an optimally structured, multigroup Production Library. Using these data and input descriptions of cells or regions, neutron spectra are generated and few-group constants are computed and fitted as a function of fuel burnup, initial enrichment, temperature, etc. In nuclear design calculations, few-group cross-section fits and descriptions of each core region and core geometry are input to a diffusion-depletion program or a nodal program that computes core reactivity, core power distribution, control rod worth, fuel cycle studies, core operating limitations, etc

  4. Advanced Transport Operating System (ATOPS) Flight Management/Flight Controls (FM/FC) software description

    Science.gov (United States)

    Wolverton, David A.; Dickson, Richard W.; Clinedinst, Winston C.; Slominski, Christopher J.

    1993-01-01

    The flight software developed for the Flight Management/Flight Controls (FM/FC) MicroVAX computer used on the Transport Systems Research Vehicle for Advanced Transport Operating Systems (ATOPS) research is described. The FM/FC software computes navigation position estimates, guidance commands, and those commands issued to the control surfaces to direct the aircraft in flight. Various modes of flight are provided for, ranging from computer assisted manual modes to fully automatic modes including automatic landing. A high-level system overview as well as a description of each software module comprising the system is provided. Digital systems diagrams are included for each major flight control component and selected flight management functions.

  5. Delayed phenomena analysis from French PWR containment instrumentation system

    International Nuclear Information System (INIS)

    Costaz, J.L.

    1987-01-01

    The analysis of the large amount of measurements which has been now gathered by EDF on its twenty two PWR 900 MW shows that the behaviour of concrete under creep and shrinkage effects is in good agreement with the values given as correct estimates by french regulations and taken into account for the design of nuclear prestressed structures. None of the containment buildings studied here showed significant differences with the regulations theoretical values and consequently all the measurements remain in the field of the allowable strain variations used for design. On the other hand, if the instant loading elastic modulus is clearly determined for each containment, and its effect on theoretical creep taken into account, it was not possible up till now to extract from measurements some particular effects such as type of concrete and agregates or climatic effects. (orig.)

  6. Reliability and reproducibility of subaxial cervical injury description system: a standardized nomenclature schema.

    Science.gov (United States)

    Bono, Christopher M; Schoenfeld, Andrew; Gupta, Giri; Harrop, James S; Anderson, Paul; Patel, Alpesh A; Dimar, John; Aarabi, Bizhan; Dailey, Andrew; Vaccaro, Alexander R; Gahr, Ralf; Shaffrey, Christopher; Anderson, David G; Rampersaud, Raj

    2011-08-01

    Radiographic measurement study. To develop a standardized cervical injury nomenclature system to facilitate description, communication, and classification among health care providers. The reliability and reproducibility of this system was then examined. Description of subaxial cervical injuries is critical for treatment decision making and comparing scientific reports of outcomes. Despite a number of available classification systems, surgeons, and researchers continue to use descriptive nomenclature, such as "burst" and "teardrop" fractures, to describe injuries. However, there is considerable inconsistency with use of such terms in the literature. Eleven distinct injury types and associated definitions were established for the subaxial cervical spine and subsequently refined by members of the Spine Trauma Study Group. A series of 18 cases of patients with a broad spectrum of subaxial cervical spine injuries was prepared and distributed to surgeon raters. Each rater was provided with the full nomenclature document and asked to select primary and secondary injury types for each case. After receipt of the raters' first round of classifications, the cases were resorted and returned to the raters for a second round of review. Interrater and intrarater reliabilities were calculated as percent agreement and Cohen kappa (κ) values. Intrarater reliability was assessed by comparing a given rater's diagnosis from the first and second rounds. Nineteen surgeons completed the first and second rounds of the study. Overall, the system demonstrated 56.4% interrater agreement and 72.8% intrarater agreement. Overall, interrater κ values demonstrated moderate agreement while intrarater κ values showed substantial agreement. Analyzed by injury types, only four (burst fractures, lateral mass fractures, flexion teardrop fractures, and anterior distraction injuries) demonstrated greater than 50% interrater agreement. This study demonstrated that, even in ideal circumstances, there is

  7. Wave kinematics and response of slender offshore structures. Vol 3: Description of measuring systems

    Energy Technology Data Exchange (ETDEWEB)

    Riber, H.J.

    1999-08-01

    The report presents the measuring systems used during the measurements in the North Sea at the Tyra field. The report consists of two parts: (1) Description of the wave measuring systems (WMS); (2) Description of the load mesuring system (LMS). The developed Wave Kinematics Measuring Systems (WMS) is an acoustic system based on the pulsed incoherent Doppler technique. Basically it consists of 4 sonar stations and a control and monitoring station. The sonar stations are positioned at the seabed north of the TCP-A platform in a star configuration with one station surrounded by the other three stations in a fixed distance of 37 m. From the central sonar stations an umbilical is routed to the TCP-A platform where the control and monitoring stations is positioned in the satellite communication room. The measurement periods have been concentrated to periods with storms. It was a special emphasis to measure the wave kinematics of extremely high waves. The LMS is described in terms of geometrical data, structural properties, instrumentation, and other characteristics. (LN)

  8. Depressurization-filtration system of the containment of French PWR's

    International Nuclear Information System (INIS)

    L'homme, A.; Schektman, N.

    1987-01-01

    In the hypothetical event of a core meltdown occurring in a pressurized water reactor, and in order to preserve the integrity of the containment threatened by a build-up in pressure, EDF has developed, with the CEA, a decompression device which filters the containment internal atmosphere by using an unused containment penetration, and a sand-box, as filtering mechanism. This device and its procedure for utilization, constitute the U5 procedure. Check-tests on a semi-industrial scale have been carried out at the Nuclear Research Centre at Cadarache, by using columns of sand 80 cm high, according to following varying criteria: the granulometry of the sand, that of the aerosols, the flow-through speed, and the percentage steam content of the fluid to be filtered. The filtering material chosen is sand of a median diameter of 0.6 mm. (log normal distribution). The purification factor is above 10. The device tested meets the chosen targets, and is applied today to French units on condition to simple modifications concerning specific aspects of different series. The first is expected to be put into service during 1987

  9. Components of formalized description of selecting tools for ensuring stability of banking system

    Directory of Open Access Journals (Sweden)

    N.P. Pogorelenko

    2015-09-01

    Full Text Available A banking system is one of the key elements of a financial market of any country. Effectiveness and functional orientation of a banking system provide continuous and targeted financial resources flowing between different sectors of economy and this allows to perform economic activities of various entities. Thus, a banking system plays an essential role in the formation of market relations. A question of stable functioning of a banking system can be defined as a key one. The basic task is to improve the management of a banking system by achieving its stability. The disclosure of formalized description of the definition of tools to influence the selection procedure for ensuring stability of a banking system should be determined as the primary objective. For reaching the goal a comparative study has been introduced and generalized concerning the concept definition of «banking system» and the ideology of its management according to the circumstances and factors of influence. The combination of individual components is to determine the instruments of influence on the banking system activity in the form of chain ties. On the base of the analysis carried out the article grounds the necessity of the generalized use of formalized description of the procedures for selecting instruments for ensuring stability of a banking system. For the purpose of this procedure the author has also grounded, determined and disclosed some of its components. To implement the relevant qualitative phase of formalization the author has proposed the use of chain patterns, and to quantify the individual parameters of such a procedure the methodology of border stochastic analysis has been offered. As a scientific novelty of the present research it is necessary to note the qualitative and quantitative phases for formal presentation of describing procedures for the selection of tools to ensure banking system stability as well as the introduction of chain schemes for the

  10. Computer system design description for SY-101 hydrogen mitigation test project data acquisition and control system (DACS-1). Revision 1

    International Nuclear Information System (INIS)

    Truitt, R.W.

    1994-01-01

    This document provides descriptions of components and tasks that are involved in the computer system for the data acquisition and control of the mitigation tests conducted on waste tank SY-101 at the Hanford Nuclear Reservation. The system was designed and implemented by Los alamos National Laboratory and supplied to Westinghouse Hanford Company. The computers (both personal computers and specialized data-taking computers) and the software programs of the system will hereafter collectively be referred to as the DACS (Data Acquisition and Control System)

  11. System design description for the SY-101 hydrogen mitigation test project data acquisition and control system (DACS-1)

    International Nuclear Information System (INIS)

    Ermi, A.M.

    1998-01-01

    There is no new activity or procedure associated with the updating of this reference document. The updating of this system design description maintains an agreed upon documentation program initiated within the test program and carried into operations at time of turnover to maintain configuration control as outlined by design authority practicing guidelines. Any changes made to controlled components in the field will be updated after the time of implementation to support the engineers and operators understand, maintain, train to and operate the system. There are no new credible failure modes associated with the updating of information in a support description document. The failure analysis of each change was reviewed at the time of implementation of the Systems Change Request for all the processes changed. This document simply provides a history of implementation and current system status. The incorporation of the two documents, Computer Systems Design Description (HNF-SD-WMCSDD-008) and the Input/Output Channel List (HNF-SD-WM-EL-001), as appendices allow for fewer errors in changes. Because the documents are all together, they will be approved as one document, not three separate entities which could be updated at different times, creating a situation which does not accurately depict field conditions

  12. Smart container UWB sensor system for situational awareness of intrusion alarms

    Science.gov (United States)

    Romero, Carlos E.; Haugen, Peter C.; Zumstein, James M.; Leach, Jr., Richard R.; Vigars, Mark L.

    2013-06-11

    An in-container monitoring sensor system is based on an UWB radar intrusion detector positioned in a container and having a range gate set to the farthest wall of the container from the detector. Multipath reflections within the container make every point on or in the container appear to be at the range gate, allowing intrusion detection anywhere in the container. The system also includes other sensors to provide false alarm discrimination, and may include other sensors to monitor other parameters, e.g. radiation. The sensor system also includes a control subsystem for controlling system operation. Communications and information extraction capability may also be included. A method of detecting intrusion into a container uses UWB radar, and may also include false alarm discrimination. A secure container has an UWB based monitoring system

  13. Use of process indices for simplification of the description of vapor deposition systems

    International Nuclear Information System (INIS)

    Kajikawa, Yuya; Noda, Suguru; Komiyama, Hiroshi

    2004-01-01

    Vapor deposition is a complex process, including gas-phase, surface, and solid-phase phenomena. Because of the complexity of chemical and physical processes occurring in vapor deposition processes, it is difficult to form a comprehensive, fundamental understanding of vapor deposition and to control such systems for obtaining desirable structures and performance. To overcome this difficulty, we present a method for simplifying the complex description of such systems. One simplification method is to separate complex systems into multiple elements, and determine which of these are important elements. We call this method abridgement. The abridgement method retains only the dominant processes in a description of the system, and discards the others. Abridgement can be achieved by using process indices to evaluate the relative importance of the elementary processes. We describe the formulation and use of these process indices through examples of the growth of continuous films, initial deposition processes, and the formation of the preferred orientation of polycrystalline films. In this paper, we propose a method for representing complex vapor deposition processes as a set of simpler processes

  14. Nuclear computerized library for assessing reactor reliability (NUCLARR): User's guide: Part 3, NUCLARR system description

    International Nuclear Information System (INIS)

    Gilmore, W.E.; Gentillon, C.D.; Gertman, D.I.; Beers, G.H.; Galyean, W.J.; Gilbert, B.G.

    1988-06-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal micro-computer. NUCLARR can be used by the end user to furnish data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume IV of this series is the User's Guide for operating the NUCLARR software and is presented in three parts. This document, Part 3: NUCLARR System Description, provides an in-depth discussion of the design characteristics and special features of the NUCLARR software. Part 3 also presents the organization of the data base structures and techniques used to manipulate the data

  15. Cold Vacuum Dryer (CVD) Facility Fire Protection System Design Description (SYS 24)

    Energy Technology Data Exchange (ETDEWEB)

    SINGH, G.

    2000-10-17

    This system design description (SDD) addresses the Cold Vacuum Drying (CVD) Facility fire protection system (FPS). The primary features of the FPS for the CVD are a fire alarm and detection system, automatic sprinklers, and fire hydrants. The FPS also includes fire extinguishers located throughout the facility and fire hydrants to assist in manual firefighting efforts. In addition, a fire barrier separates the operations support (administrative) area from the process bays and process bay support areas. Administrative controls to limit combustible materials have been established and are a part of the overall fire protection program. The FPS is augmented by assistance from the Hanford Fire Department (HED) and by interface systems including service water, electrical power, drains, instrumentation and controls. This SDD, when used in conjunction with the other elements of the definitive design package, provides a complete picture of the FPS for the CVD Facility.

  16. MPEG-7-based description infrastructure for an audiovisual content analysis and retrieval system

    Science.gov (United States)

    Bailer, Werner; Schallauer, Peter; Hausenblas, Michael; Thallinger, Georg

    2005-01-01

    We present a case study of establishing a description infrastructure for an audiovisual content-analysis and retrieval system. The description infrastructure consists of an internal metadata model and access tool for using it. Based on an analysis of requirements, we have selected, out of a set of candidates, MPEG-7 as the basis of our metadata model. The openness and generality of MPEG-7 allow using it in broad range of applications, but increase complexity and hinder interoperability. Profiling has been proposed as a solution, with the focus on selecting and constraining description tools. Semantic constraints are currently only described in textual form. Conformance in terms of semantics can thus not be evaluated automatically and mappings between different profiles can only be defined manually. As a solution, we propose an approach to formalize the semantic constraints of an MPEG-7 profile using a formal vocabulary expressed in OWL, which allows automated processing of semantic constraints. We have defined the Detailed Audiovisual Profile as the profile to be used in our metadata model and we show how some of the semantic constraints of this profile can be formulated using ontologies. To work practically with the metadata model, we have implemented a MPEG-7 library and a client/server document access infrastructure.

  17. Fascia: a morphological description and classification system based on a literature review

    Science.gov (United States)

    Kumka, Myroslava; Bonar, Jason

    2012-01-01

    Fascia is virtually inseparable from all structures in the body and acts to create continuity amongst tissues to enhance function and support. In the past fascia has been difficult to study leading to ambiguities in nomenclature, which have only recently been addressed. Through review of the available literature, advances in fascia research were compiled, and issues related to terminology, descriptions, and clinical relevance of fascia were addressed. Our multimodal search strategy was conducted in Medline and PubMed databases, with other targeted searches in Google Scholar and by hand, utilizing reference lists and conference proceedings. In an effort to organize nomenclature for fascial structures provided by the Federative International Committee on Anatomical Terminology (FICAT), we developed a functional classification system which includes four categories of fascia: i) linking, ii) fascicular, iii) compression, and iv) separating fasciae. Each category was developed from descriptions in the literature on gross anatomy, histology, and biomechanics; the category names reflect the function of the fascia. An up-to-date definition of fascia is provided, as well as descriptions of its function and clinical features. Our classification demonstrates the use of internationally accepted terminology in an ontology which can improve understanding of major terms in each category of fascia. PMID:22997468

  18. Uranium accountability for ATR fuel fabrication. Part I. A description of the existing system

    International Nuclear Information System (INIS)

    Dolan, C.A.; Nieschmidt, E.B.; Vegors, S.H. Jr.; Wagner, E.P. Jr.

    1977-06-01

    An evaluation of the materials accountability program at the Atomics International fuel fabrication facility in Canoga Park, California, with regard to the fabrication of highly enriched uranium fuel for the Advanced Test Reactor is presented. An analysis is given of the existing standards program, the existing measurements program and the existing statistical analysis procedures. In addition a short discussion is given of our evaluation of the safeguards procedures at Atomics International together with suggestions for possible modifications and improvements. Appendices of this report contain a rather complete description of the Atomics International plant and the flow of highly enriched uranium through the plant as well as the principal documents used for material accountability records

  19. Development of the interactive model between Component Cooling Water System and Containment Cooling System using GOTHIC

    International Nuclear Information System (INIS)

    Byun, Choong Sup; Song, Dong Soo; Jun, Hwang Yong

    2006-01-01

    In a design point of view, component cooling water (CCW) system is not full-interactively designed with its heat loads. Heat loads are calculated from the CCW design flow and temperature condition which is determined with conservatism. Then the CCW heat exchanger is sized by using total maximized heat loads from above calculation. This approach does not give the optimized performance results and the exact trends of CCW system and the loads during transient. Therefore a combined model for performance analysis of containment and the component cooling water(CCW) system is developed by using GOTHIC software code. The model is verified by using the design parameters of component cooling water heat exchanger and the heat loads during the recirculation mode of loss of coolant accident scenario. This model may be used for calculating the realistic containment response and CCW performance, and increasing the ultimate heat sink temperature limits

  20. Binding in some few-body systems containing antimatter

    International Nuclear Information System (INIS)

    Armour, E.A.G.

    2009-01-01

    It is well known that the system made up of a fixed proton and antiproton and an electron (or a positron) has no bound states if the internuclear distance R 0 . In this paper, I consider the more complicated system in which the electron and the positron are both present and investigate the possibility of obtaining a lower bound on the value of R for which the system has no bound states. I also investigate the implications of the existence of bound states of the simpler, one light particle system regarding bound states of the more complicated system. This article is based on the presentation by E. A. G. Armour at the Fifth Workshop on Critical Stability, Erice, Sicily. (author)

  1. Aqueous-salt system containing ytterbium nitrate and pyridine nitrate

    International Nuclear Information System (INIS)

    Zhuravlev, E.F.; Khisaeva, D.A.; Izmajlova, L.V.

    1983-01-01

    Cross-section method has been used to study solubility in ternary aqueous-salt system Yb(NO 3 ) 3 -C 5 H 5 NxHNO 3 -H 2 0 at 25 and 50 deg C. It is established that the system is characterized by chemical interaction. Congruently soluble compound of Yb(NO 3 ) 3 x2[C 5 H 5 NxHNO 3 ] composition is discovered in the system. Composition of the compound is confirmed by chemical analysis; its infrared spectra are studied. Interplanar distances are determined; derivatogram of the compound is given. The results of the works are compared with analogous investigations of another rare earth nitrates

  2. Tritium containment in fusion facilities

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1978-01-01

    The key environmental control systems that have been identified and are being developed are listed. A brief description of each of the following systems is given: primary process materials, permeation barriers, secondary containment, tritium waste treatment, emergency tritium cleanup, maintenance procedures, and tertiary containment

  3. 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

  4. System requirements and design description for the document basis database interface (DocBasis)

    International Nuclear Information System (INIS)

    Lehman, W.J.

    1997-01-01

    This document describes system requirements and the design description for the Document Basis Database Interface (DocBasis). The DocBasis application is used to manage procedures used within the tank farms. The application maintains information in a small database to track the document basis for a procedure, as well as the current version/modification level and the basis for the procedure. The basis for each procedure is substantiated by Administrative, Technical, Procedural, and Regulatory requirements. The DocBasis user interface was developed by Science Applications International Corporation (SAIC)

  5. Unambiguous discrimination of mixed states: A description based on system-ancilla coupling

    International Nuclear Information System (INIS)

    Zhou, Xiang-Fa; Zhang, Yong-Sheng; Guo, Guang-Can

    2007-01-01

    We propose a general description for the unambiguous discrimination of mixed states according to the system-environment coupling, and present a procedure to reduce this to a standard semidefinite programming problem. In the two-state case, we introduce the canonical vectors and partly simplify the problem to the case of discrimination between pairs of canonical vectors. By considering the positivity of the 2x2 matrices, we obtain a series of new upper bounds for the total success probability, which depends on both the prior probabilities and specific state structures

  6. Reward Systems and Performance of Sales: A Descriptive Study among the Ghanaian Insurance Industry

    Directory of Open Access Journals (Sweden)

    Joshua Ohene-Danso

    2015-04-01

    Full Text Available Ghanaian managers over recent years have taken a steady pattern of organizational policies, aimed specifically at enhancing employees’ development and management. Significant among these measures are recognition and rewards management. The system of rewards at selected Ghanaian Insurance Companies within it Southern Sector operations affected the performance of employees in the sales and marketing of products. Descriptive results indicate that, reward strategies are significant in providing an incentive to employees to work. It is recommended that total rewards should be extended to cover job security and other benefits in the form of recognition.

  7. Background and system description of the Mod 1 wind turbine generator

    Science.gov (United States)

    Ernst, E. H.

    1978-01-01

    The Mod-1 wind turbine considered is a large utility-class machine, operating in the high wind regime, which has the potential for generation of utility grade power at costs competitive with other alternative energy sources. A Mod-1 wind turbine generator (WTG) description is presented, taking into account the two variable-pitch steel blades of the rotor, the drive train, power generation/control, the Nacelle structure, and the yaw drive. The major surface elements of the WTG are the ground enclosure, the back-up battery system, the step-up transformer, elements of the data system, cabling, area lighting, and tower foundation. The final system weight (rotor, Nacelle, and tower) is expected to be about 650,000 pounds. The WTG will be capable of delivering 1800 kW to the utility grid in a wind-speed above 25 mph.

  8. The achievement and assessment of safety in systems containing software

    International Nuclear Information System (INIS)

    Ball, A.; Dale, C.J.; Butterfield, M.H.

    1986-01-01

    In order to establish confidence in the safe operation of a reactor protection system, there is a need to establish, as far as it is possible, that: (i) the algorithms used are correct; (ii) the system is a correct implementation of the algorithms; and (iii) the hardware is sufficiently reliable. This paper concentrates principally on the second of these, as it applies to the software aspect of the more accurate and complex trip functions to be performed by modern reactor protection systems. In order to engineer safety into software, there is a need to use a development strategy which will stand a high chance of achieving a correct implementation of the trip algorithms. This paper describes three broad methodologies by which it is possible to enhance the integrity of software: fault avoidance, fault tolerance and fault removal. Fault avoidance is concerned with making the software as fault free as possible by appropriate choice of specification, design and implementation methods. A fault tolerant strategy may be advisable in many safety critical applications, in order to guard against residual faults present in the software of the installed system. Fault detection and removal techniques are used to remove as many faults as possible of those introduced during software development. The paper also discusses safety and reliability assessment as it applies to software, outlining the various approaches available. Finally, there is an outline of a research project underway in the UKAEA which is intended to assess methods for developing and testing safety and protection systems involving software. (author)

  9. A self-description data framework for Tokamak control system design

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Ming; Zhang, Jing [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zheng, Wei, E-mail: zhengwei@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Hu, Feiran; Zhuang, Ge [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2015-10-15

    Highlights: • The SDD framework can be applied to different Tokamak devices. • We explain how configuration settings of control systems are described in SDD models, namely components and connections. • Evolving SDD models are stored in a dynamic schema database. • The SDD editor supports plug-and-play SDD models. - Abstract: A Tokamak device consists of numerous control systems, which need to be integrated. CODAC (Control, Data Access and Communication) system requires the configuration settings of these control systems to carry out the integration smoothly. SDD (Self-description data) is designed to describe the static configuration of control systems. ITER CODAC group has released an SDD software package for control system designers to manage the static configuration, but it is specific for ITER plant control systems. Following the idea of ITER SDD, we developed a flexible and scalable SDD framework to develop SDD software for J-TEXT and other sophisticated devices. The SDD framework describes the configuration settings of various control systems, including physical and logical elements and their relation information, in SDD models which are classified into Components and Connections. The framework is composed of three layers: the MongoDB database, an open-source, dynamic schema, NoSQL (Not Only SQL) database; the SDD service, which maps SDD models to MongoDB and handles the transaction and business logic; the SDD applications, which can be used to create and maintain SDD information, and generate various kinds of output using the stored SDD information.

  10. A self-description data framework for Tokamak control system design

    International Nuclear Information System (INIS)

    Zhang, Ming; Zhang, Jing; Zheng, Wei; Hu, Feiran; Zhuang, Ge

    2015-01-01

    Highlights: • The SDD framework can be applied to different Tokamak devices. • We explain how configuration settings of control systems are described in SDD models, namely components and connections. • Evolving SDD models are stored in a dynamic schema database. • The SDD editor supports plug-and-play SDD models. - Abstract: A Tokamak device consists of numerous control systems, which need to be integrated. CODAC (Control, Data Access and Communication) system requires the configuration settings of these control systems to carry out the integration smoothly. SDD (Self-description data) is designed to describe the static configuration of control systems. ITER CODAC group has released an SDD software package for control system designers to manage the static configuration, but it is specific for ITER plant control systems. Following the idea of ITER SDD, we developed a flexible and scalable SDD framework to develop SDD software for J-TEXT and other sophisticated devices. The SDD framework describes the configuration settings of various control systems, including physical and logical elements and their relation information, in SDD models which are classified into Components and Connections. The framework is composed of three layers: the MongoDB database, an open-source, dynamic schema, NoSQL (Not Only SQL) database; the SDD service, which maps SDD models to MongoDB and handles the transaction and business logic; the SDD applications, which can be used to create and maintain SDD information, and generate various kinds of output using the stored SDD information.

  11. Correction of chromatic abberation in electrostatic lense systems containing quadrupoles

    International Nuclear Information System (INIS)

    Baranova, L.A.; Ul'yanova, N.S.; Yavor, S.Ya.

    1991-01-01

    Possibility of chromatic abberation correction in immersion systems consisting of axysimmetric and quadrupole lenses is shown. Concrete examples are presented. A number of new directions in science and technique, using ion beams are intensively developed presently. When using them accute necessity arises in chromatic abberation correction, while large-scale energy scattering is observed as a rule in such cases

  12. Tracer verification and monitoring of containment systems (II)

    International Nuclear Information System (INIS)

    Williams, C.V.; Dunn, S.D.; Lowry, W.E.

    1997-01-01

    A tracer verification and monitoring system, SEAtrace trademark, has been designed and field tested which uses gas tracers to evaluate, verify, and monitor the integrity of subsurface barriers. This is accomplished using an automatic, rugged, autonomous monitoring system combined with an inverse optimization code. A gaseous tracer is injected inside the barrier and an array of wells outside the barrier are monitored. When the tracer gas is detected, a global optimization code is used to calculate the leak parameters, including leak size, location, and when the leak began. The multipoint monitoring system operates in real-time, can be used to measure both the tracer gas and soil vapor contaminants, and is capable of unattended operation for long periods of time (months). The global optimization code searches multi-dimensional open-quotes spaceclose quotes to find the best fit for all of the input parameters. These parameters include tracer gas concentration histories from multiple monitoring points, medium properties, barrier location, and the source concentration. SEAtrace trademark does not attempt to model all of the nuances associated with multi-phase, multi-component flow, but rather, the inverse code uses a simplistic forward model which can provide results which are reasonably accurate. The system has calculated leak locations to within 0.5 meters and leak radii to within 0.12 meters

  13. V1334 Cyg: A Triple System Containing a Classical Cepheid

    Science.gov (United States)

    Evans, N. R.

    2000-05-01

    HR 8157 = ADS 14859 = HD 203156 = V1334 Cyg was recognized a hundred years ago to be a marginally resolved visual binary. Millis (1969, Lowell Obs Bull, 7, 113) discovered that the brightest star in the system is a low amplitude classical Cepheid with a pulsation period of 3.3 days. Early radial velocity observations by Abt and Levy (1970, PASP, 82, 334) differed from scattered radial velocity observations in the first half of the century implying that in addition to the long period system, the Cepheid is also a member of a short period binary. We have observed Cepheid V1334 Cyg A for nearly 30 years. From this radial velocity data we have derived an orbit with a period of 5 years. The orbit provides limits on the mass of the companion (V1334 Cyg C) of 3.1 to 4.4 solar masses. We have used an IUE high resolution spectrum to conclude that the hottest star in the system (V1334 Cyg B) which dominates the spectrum in the ultraviolet is the wide companion since the velocity is very near the systemic velocity. Financial support was supplied through a Natural Sciences and Engineering Research Council, Canada (NSERC) grant and HST Grant GO-07478.01-96A, and from the Chandra Science Center NASA Contract NAS8-39073.

  14. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  15. Source term aspects associated with future PWR containment systems

    International Nuclear Information System (INIS)

    Kuczera, B.; Kebler, G.; Ehrhardt, J.; Scholtyssek, W.

    1994-01-01

    The overall objective of reactor safety is to protect the population against dangerous releases of radioactive materials from nuclear power plants. In context with a reinforcement of the defense-in-depth strategy the common safety requirements on future nuclear power plants converge in the objective that these plants should be so safe that even in case of a severe accident there will be no need of off-site emergency actions such as an evacuation or resettlement of the population from the vicinity of a nuclear power plant. It is shown by the example of a future 1400 MWe pressurized water reactor (PWR) plant that this goal can be attained in principle by providing a double containment with the annulus vented via an appropriate emergency standby filter. Within the framework of severe accident consequence mitigation a set of parameters for accident conditions and emergency filter efficiencies is elaborated under which the German lower levels of intervention for evacuation are not attained. (author). 10 refs., 3 tabs., 5 figs

  16. Application of agent-based system for bioprocess description and process improvement.

    Science.gov (United States)

    Gao, Ying; Kipling, Katie; Glassey, Jarka; Willis, Mark; Montague, Gary; Zhou, Yuhong; Titchener-Hooker, Nigel J

    2010-01-01

    Modeling plays an important role in bioprocess development for design and scale-up. Predictive models can also be used in biopharmaceutical manufacturing to assist decision-making either to maintain process consistency or to identify optimal operating conditions. To predict the whole bioprocess performance, the strong interactions present in a processing sequence must be adequately modeled. Traditionally, bioprocess modeling considers process units separately, which makes it difficult to capture the interactions between units. In this work, a systematic framework is developed to analyze the bioprocesses based on a whole process understanding and considering the interactions between process operations. An agent-based approach is adopted to provide a flexible infrastructure for the necessary integration of process models. This enables the prediction of overall process behavior, which can then be applied during process development or once manufacturing has commenced, in both cases leading to the capacity for fast evaluation of process improvement options. The multi-agent system comprises a process knowledge base, process models, and a group of functional agents. In this system, agent components co-operate with each other in performing their tasks. These include the description of the whole process behavior, evaluating process operating conditions, monitoring of the operating processes, predicting critical process performance, and providing guidance to decision-making when coping with process deviations. During process development, the system can be used to evaluate the design space for process operation. During manufacture, the system can be applied to identify abnormal process operation events and then to provide suggestions as to how best to cope with the deviations. In all cases, the function of the system is to ensure an efficient manufacturing process. The implementation of the agent-based approach is illustrated via selected application scenarios, which

  17. Design of Drug Delivery Systems Containing Artemisinin and Its Derivatives

    Directory of Open Access Journals (Sweden)

    Blessing Atim Aderibigbe

    2017-02-01

    Full Text Available Artemisinin and its derivatives have been reported to be experimentally effective for the treatment of highly aggressive cancers without developing drug resistance, they are useful for the treatment of malaria, other protozoal infections and they exhibit antiviral activity. However, they are limited pharmacologically by their poor bioavailability, short half-life in vivo, poor water solubility and long term usage results in toxicity. They are also expensive for the treatment of malaria when compared to other antimalarials. In order to enhance their therapeutic efficacy, they are incorporated onto different drug delivery systems, thus yielding improved biological outcomes. This review article is focused on the currently synthesized derivatives of artemisinin and different delivery systems used for the incorporation of artemisinin and its derivatives.

  18. Validation of NCSSHP for highly enriched uranium systems containing beryllium

    International Nuclear Information System (INIS)

    Krass, A.W.; Elliott, E.P.; Tollefson, D.A.

    1994-01-01

    This document describes the validation of KENO V.a using the 27-group ENDF/B-IV cross section library for highly enriched uranium and beryllium neutronic systems, and is in accordance with ANSI/ANS-8.1-1983(R1988) requirements for calculational methods. The validation has been performed on a Hewlett Packard 9000/Series 700 Workstation at the Oak Ridge Y-12 Plant Nuclear Criticality Safety Department using the Oak Ridge Y-12 Plant Nuclear Criticality Safety Software code package. Critical experiments from LA-2203, UCRL-4975, ORNL-2201, and ORNL/ENG-2 have been identified as having the constituents desired for this validation as well as sufficient experimental detail to allow accurate construction of KENO V.a calculational models. The results of these calculations establish the safety criteria to be employed in future calculational studies of these types of systems

  19. System design description for mini-dacs data acquisition and control system

    International Nuclear Information System (INIS)

    Vargo, F.G. Jr.; Trujillo, L.T.; Smith, S.O.

    1994-01-01

    This document describes the hardware computer system, for the mini data acquisition and control system (DACS) that was fabricated by Los Alamos National Laboratory (LANL), to support the testing of the spare mixer pump for SY-101

  20. Nuclear containment systems and in-service inspection status of Korea nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jihong, Park; Jaekeun, Hong; Banuk, Park [Korea Institute of Machinery and Materials, Dept. of Authorized Test and Evaluation, Kyungnam (Korea, Republic of)

    2007-07-01

    20 unit nuclear power plants in Korea have been operated and maintained since the first unit started in commercial service in 1978. Most recently 4 units were under construction and several units were planned to be constructed. by industries. 4 types of nuclear containment systems have been constructed until now: first, metal containments, then pre-stressed concrete containments with grouted tendon systems, followed by pre-stressed concrete containments with un-grouted tendon systems, and Korea standard nuclear containments. All the nuclear containments should be inspected periodically. Therefore for periodic in-service inspection, several appropriate technical requirements should be applied differently depending on the specific nuclear containment types. With the changes of times, nuclear containment systems have undergone a remarkable change, and finally nuclear containment system of Korea standard nuclear power plant was settled down, and as a matter of course it dominates the trend of present and future nuclear containment systems. Overall in-service inspection results of most Korea nuclear containments have not showed any serious evidence of degradation.