WorldWideScience

Sample records for design basis temperatures

  1. Selection of design basis event for modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

    2016-06-01

    Japan Atomic Energy Agency (JAEA) has been investigating safety requirements and basic approach of safety guidelines for modular High Temperature Gas-cooled Reactor (HTGR) aiming to increase internarial contribution for nuclear safety by developing an international HTGR safety standard under International Atomic Energy Agency. In this study, we investigate a deterministic approach to select design basis events utilizing information obtained from probabilistic approach. In addition, selections of design basis events are conducted for commercial HTGR designed by JAEA. As a result, an approach for selecting design basis event considering multiple failures of safety systems is established which has not been considered as design basis in the safety guideline for existing nuclear facility. Furthermore, selection of design basis events for commercial HTGR has completed. This report provides an approach and procedure for selecting design basis events of modular HTGR as well as selected events for the commercial HTGR, GTHTR300. (author)

  2. Creep-Fatigue Life Design with Various Stress and Temperature Conditions on the Basis of Lethargy Coefficient

    International Nuclear Information System (INIS)

    Park, Jung Eun; Yang, Sung Mo; Han, Jae Hee; Yu, Hyo Sun

    2011-01-01

    High temperature and stress are encounted in power plants and vehicle engines. Therefore, determination of the creep-fatigue life of a material is necessary prior to fabricating equipment. In this study, life design was determined on the basis of the lethargy coefficient for different temperatures, stress and rupture times. SP-Creep test data was compared with computed data. The SP-Creep test was performed to obtain the rupture time for X20CrMoV121 steel. The integration life equation was considered for three cases with various load, temperature and load-temperature. First, the lethargy coefficient was calculated by using the obtained rupture stress and the rupture time that were determined by carrying out the SP-Creep test. Next, life was predicted on the basis of the temperature condition. Finally, it was observed that life decreases considerably due to the coupling effect that results when fatigue and creep occur simultaneously

  3. Design basis 2

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, G.; Soerensen, P. [Risoe National Lab., Roskilde (Denmark)

    1996-09-01

    Design Basis Program 2 (DBP2) is comprehensive fully coupled code which has the capability to operate in the time domain as well as in the frequency domain. The code was developed during the period 1991-93 and succeed Design Basis 1, which is a one-blade model presuming stiff tower, transmission system and hub. The package is designed for use on a personal computer and offers a user-friendly environment based on menu-driven editing and control facilities, and with graphics used extensively for the data presentation. Moreover in-data as well as results are dumped on files in Ascii-format. The input data is organized in a in-data base with a structure that easily allows for arbitrary combinations of defined structural components and load cases. (au)

  4. Interior design conceptual basis

    CERN Document Server

    Sully, Anthony

    2015-01-01

    Maximizing reader insights into interior design as a conceptual way of thinking, which is about ideas and how they are formulated. The major themes of this book are the seven concepts of planning, circulation, 3D, construction, materials, colour and lighting, which covers the entire spectrum of a designer’s activity. Analysing design concepts from the view of the range of possibilities that the designer can examine and eventually decide by choice and conclusive belief the appropriate course of action to take in forming that particular concept, the formation and implementation of these concepts is taken in this book to aid the designer in his/her professional task of completing a design proposal to the client. The purpose of this book is to prepare designers to focus on each concept independently as much as possible, whilst acknowledging relative connections without unwarranted influences unfairly dictating a conceptual bias, and is about that part of the design process called conceptual analysis. It is assu...

  5. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    In 1985 an incident at Toledo Edison's Davis Besse plant caused the U.S. Nuclear Regulatory Commission (NRC) to re-evaluate the technical information that the utilities had readily available to support the design of their plants. The Design Basis programs, currently on going in most U.S. utilities, have been the nuclear industry's response to the needs identified by this re-evaluation. In order to understand the Design Basis programs which have been implemented by the U.S. nuclear utilities, it is necessary to understand the problem as it was perceived by the nuclear industry (the utilities, the original NSSS designers and the regulators) after the Davis-Besse incident, the subsequent programs undertaken by the industry under the leadership of INPO and NUMARC, the NRC's actions, and the overall evolution of the industry's vision in relation to this problem. This paper presents the history of the design basis efforts from the first recognition of the problem by the NRC after the Davis-Besse incident, describes the actions taken by the NRC, INPO, NUMARC, the U.S. utilities and the NSSS designers, and brings the problem statement up-to-date in relation to the vision presently held by the U.S. nuclear industry. It then presents a technical discussion to develop a detailed definition of design basis information to support the problem statement. The information originally supplied by the NSSS designers during the plant design and construction is discussed as well as its relationship to the previously defined design basis information. This section of the paper concludes by defining the additional information needed by nuclear utilities to satisfy the requirements developed from the problem statement. Having developed a definition of the additional information (i.e., information not originally supplied during design and construction) required to solve the design basis problem as it is presently perceived by the U.S. nuclear industry, the paper then discusses design basis

  6. Design basis II: Design for events

    International Nuclear Information System (INIS)

    Frisch, W.

    1982-01-01

    In a lecture of this title, it could be expected that all events which are a basis for system and component design are described. According to the title of the Course 'Instrumentation and Control of Nuclear Power Plants' emphasis is put on events originating within the plant (no consideration of external events such as air plane crash or earth-quake). The lecture is divided into the two parts 'Transients' and 'Loss of coolant accidents (LOCAs)'. Due to the complex interaction between systems and components during transients, the first part is the main part of the lecture, while the second part (LOCAs) is only a very brief description of emergency core cooling system functions and the typical course of a large and small LOCA event. The first part on anticipated transients with intact primary coolant system boundary (non-LOCA-transients) covers several aspects of the analysis, such as classification, brief system description, transient description, analysis of anticipated transients without scram (ATWS) and analytical methods. Due to the time restriction necessary within the course, only a small section of the entire area can be presented in this paper. (orig.)

  7. Understanding and capturing NSSS design basis

    International Nuclear Information System (INIS)

    Palo, W.J.; Miller, B.

    1993-01-01

    Changes to, and technical evaluations of nuclear generating station designs are often warranted. Comprehensive documentation and understanding of the NSSS Design Basis are essential to support these activities. Effective configuration management tools are also needed to maintain the plant within design basis limits. Efficient design basis reconstitution can be realized via: In-depth understanding of the design process; Utilization of effective data collection methodology; State of the art data basing tools. A database can be created to generate a Design Basis Manual (DBM). This database can communicate electronically with other plant databases. A living document vice a static snapshot of the plant design is the goal. A design basis database can serve as the cornerstone for a global electronic information control system

  8. Determination of Design Basis Earthquake ground motion

    International Nuclear Information System (INIS)

    Kato, Muneaki

    1997-01-01

    This paper describes principle of determining of Design Basis Earthquake following the Examination Guide, some examples on actual sites including earthquake sources to be considered, earthquake response spectrum and simulated seismic waves. In sppendix of this paper, furthermore, seismic safety review for N.P.P designed before publication of the Examination Guide was summarized with Check Basis Earthquake. (J.P.N.)

  9. Determination of Design Basis Earthquake ground motion

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Muneaki [Japan Atomic Power Co., Tokyo (Japan)

    1997-03-01

    This paper describes principle of determining of Design Basis Earthquake following the Examination Guide, some examples on actual sites including earthquake sources to be considered, earthquake response spectrum and simulated seismic waves. In sppendix of this paper, furthermore, seismic safety review for N.P.P designed before publication of the Examination Guide was summarized with Check Basis Earthquake. (J.P.N.)

  10. Exploratory Shaft Facility design basis study report

    International Nuclear Information System (INIS)

    Langstaff, A.L.

    1987-01-01

    The Design Basis Study is a scoping/sizing study that evaluated the items concerning the Exploratory Shaft Facility Design including design basis values for water and methane inflow; flexibility of the design to support potential changes in program direction; cost and schedule impacts that could result if the design were changed to comply with gassy mine regulations; and cost, schedule, advantages and disadvantages of a larger second shaft. Recommendations are proposed concerning water and methane inflow values, facility layout, second shaft size, ventilation, and gassy mine requirements. 75 refs., 3 figs., 7 tabs

  11. Design Characteristics as Basis for Design Languages

    DEFF Research Database (Denmark)

    Mortensen, Niels Henrik

    1997-01-01

    The application of modern feature based CAD systems has in many companies lead to significant rationalisation of design, particulary the "down stream" acticities such as NC code generation, FEM analysis, mould flow simulation and documentation. The subject of this paper is the "up stream" activit......The application of modern feature based CAD systems has in many companies lead to significant rationalisation of design, particulary the "down stream" acticities such as NC code generation, FEM analysis, mould flow simulation and documentation. The subject of this paper is the "up stream...

  12. FRG conceptual design and design basis

    International Nuclear Information System (INIS)

    Roethemeyer, H.

    1979-01-01

    For the site-independent conceptual design the following requirements have been laid down: (1) for safety reasons retrievability is not considered; (2) standard mining techniques and experience gained at Asse should be used; (3) two shafts should be sufficient; (4) different waste forms and containers shall be disposed of in different storage areas; (5) ventilated sections must allow the shutting off of each storage area from the rest of the mine; (6) the mining method of retreat working should be applied; (7) the mine works shall have a lateral safety distance to the caprock of 200 m and a vertical safety zone beneath salt level of 300 m; (8) all disposal areas shall be on one level; (9) salt and waste shall be transported in different drifts, mainly in a one way system

  13. Assessment of WWER fuel condition in design basis accident

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.; Sokolov, N.; Andreeva-Andrievskaya, L.; Vlasov, Yu.; Nechaeva, O.; Salatov, A.

    1994-01-01

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR's. 12 figs., 11 refs

  14. Assessment of WWER fuel condition in design basis accident

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu; Sokolov, N; Andreeva-Andrievskaya, L; Vlasov, Yu; Nechaeva, O; Salatov, A [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

    1994-12-31

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR`s. 12 figs., 11 refs.

  15. PRELIMINARY SELECTION OF MGR DESIGN BASIS EVENTS

    International Nuclear Information System (INIS)

    Kappes, J.A.

    1999-01-01

    The purpose of this analysis is to identify the preliminary design basis events (DBEs) for consideration in the design of the Monitored Geologic Repository (MGR). For external events and natural phenomena (e.g., earthquake), the objective is to identify those initiating events that the MGR will be designed to withstand. Design criteria will ensure that radiological release scenarios resulting from these initiating events are beyond design basis (i.e., have a scenario frequency less than once per million years). For internal (i.e., human-induced and random equipment failures) events, the objective is to identify credible event sequences that result in bounding radiological releases. These sequences will be used to establish the design basis criteria for MGR structures, systems, and components (SSCs) design basis criteria in order to prevent or mitigate radiological releases. The safety strategy presented in this analysis for preventing or mitigating DBEs is based on the preclosure safety strategy outlined in ''Strategy to Mitigate Preclosure Offsite Exposure'' (CRWMS M andO 1998f). DBE analysis is necessary to provide feedback and requirements to the design process, and also to demonstrate compliance with proposed 10 CFR 63 (Dyer 1999b) requirements. DBE analysis is also required to identify and classify the SSCs that are important to safety (ITS)

  16. System Design and the Safety Basis

    International Nuclear Information System (INIS)

    Ellingson, Darrel

    2008-01-01

    The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination and decommissioning (D and D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities

  17. Prototype Hanford Surface Barrier: Design basis document

    International Nuclear Information System (INIS)

    Myers, D.R.; Duranceau, D.A.

    1994-11-01

    The Hanford Site Surface Barrier Development Program (BDP) was organized in 1985 to develop the technology needed to provide a long-term surface barrier capability for the Hanford Site and other arid sites. This document provides the basis of the prototype barrier. Engineers and scientists have momentarily frozen evolving barrier designs and incorporated the latest findings from BDP tasks. The design and construction of the prototype barrier has required that all of the various components of the barrier be brought together into an integrated system. This integration is particularly important because some of the components of the protective barreir have been developed independently of other barreir components. This document serves as the baseline by which future modifications or other barrier designs can be compared. Also, this document contains the minutes of meeting convened during the definitive design process in which critical decisions affecting the prototype barrier's design were made and the construction drawings

  18. Establishing 'design basis threat' in Norway

    International Nuclear Information System (INIS)

    Maerli, M.B.; Naadland, E.; Reistad, O.

    2002-01-01

    Full text: INFCIRC 225 (Rev. 4) assumes that a state's physical protection system should be based on the state's evaluation of the threat, and that this should be reflected in the relevant legislation. Other factors should also be considered, including the state's emergency response capabilities and the existing and relevant measures of the state's system of accounting for and control of nuclear material. A design basis threat developed from an evaluation by the state of the threat of unauthorized removal of nuclear material and of sabotage of nuclear material and nuclear facilities is an essential element of a state's system of physical protection. The state should continuously review the threat, and evaluate the implications of any changes in that threat for the required levels and the methods of physical protection. As part of a national design basis threat assessment, this paper evaluates the risk of nuclear or radiological terrorism and sabotage in Norway. Possible scenarios are presented and plausible consequences are discussed with a view to characterize the risks. The need for more stringent regulatory requirements will be discussed, together with the (positive) impact of improved systems and procedures of physical protection on nuclear emergency planning. Special emphasis is placed on discussing the design basis threat for different scenarios in order to systemize regulatory efforts to update the current legislation, requirement for operators' contingency planning, response efforts and the need for emergency exercises. (author)

  19. Design Load Basis for Offshore Wind turbines

    DEFF Research Database (Denmark)

    Natarajan, Anand; Hansen, Morten Hartvig; Wang, Shaofeng

    2016-01-01

    DTU Wind Energy is not designing and manufacturing wind turbines and does therefore not need a Design Load Basis (DLB) that is accepted by a certification body. However, to assess the load consequences of innovative features and devices added to existing offshore turbine concepts or new offshore...... turbine concept developed in our research, it is useful to have a full DLB that follows the current design standard and is representative of a general DLB used by the industry. It will set a standard for the offshore wind turbine design load evaluations performed at DTU Wind Energy, which is aligned...... with the challenges faced by the industry and therefore ensures that our research continues to have a strong foundation in this interaction. Furthermore, the use of a full DLB that follows the current standard can improve and increase the feedback from the research at DTU Wind Energy to the international...

  20. High temperature pipeline design

    Energy Technology Data Exchange (ETDEWEB)

    Greenslade, J.G. [Colt Engineering, Calgary, AB (Canada). Pipelines Dept.; Nixon, J.F. [Nixon Geotech Ltd., Calgary, AB (Canada); Dyck, D.W. [Stress Tech Engineering Inc., Calgary, AB (Canada)

    2004-07-01

    It is impractical to transport bitumen and heavy oil by pipelines at ambient temperature unless diluents are added to reduce the viscosity. A diluted bitumen pipeline is commonly referred to as a dilbit pipeline. The diluent routinely used is natural gas condensate. Since natural gas condensate is limited in supply, it must be recovered and reused at high cost. This paper presented an alternative to the use of diluent to reduce the viscosity of heavy oil or bitumen. The following two basic design issues for a hot bitumen (hotbit) pipeline were presented: (1) modelling the restart problem, and, (2) establishing the maximum practical operating temperature. The transient behaviour during restart of a high temperature pipeline carrying viscous fluids was modelled using the concept of flow capacity. Although the design conditions were hypothetical, they could be encountered in the Athabasca oilsands. It was shown that environmental disturbances occur when the fluid is cooled during shut down because the ground temperature near the pipeline rises. This can change growing conditions, even near deeply buried insulated pipelines. Axial thermal loads also constrain the design and operation of a buried pipeline as higher operating temperatures are considered. As such, strain based design provides the opportunity to design for higher operating temperature than allowable stress based design methods. Expansion loops can partially relieve the thermal stress at a given temperature. As the design temperature increase, there is a point at which above grade pipelines become attractive options, although the materials and welding procedures must be suitable for low temperature service. 3 refs., 1 tab., 10 figs.

  1. [Basis for designing a medical course curriculum].

    Science.gov (United States)

    Villarreal, R; Bojalil, L F; Mercer, H

    1977-01-01

    This article sets forth the reasons for the structure given to the Division of Biology and Health on the Xochimilco campus of Metropolitan Autonomous University in Mexico: to adjust the university to the process of social change going forward in the country and gear the university to the problems of the present by avoiding the rigidity of its structure. The basic aspects of curriculum design are cited against a background of an historical analysis of the socioeconomic structure of education and health. The principles underlying the curriculum and the course work are then described on the basis of that analysis.

  2. Designing for elevated temperature

    International Nuclear Information System (INIS)

    Boer, G.A. de

    1982-01-01

    The reasons for the application of higher process temperatures are explained. The properties of stainless steel are compared with those of other materials such as molybdenum. Factors influencing the choice of the material such as availability of material data at high temperature, controllability, and strength of heat-affected zone are discussed. The process of designing a structure for safe and economic high-temperature application is outlined: design-by-analysis in contrast to the design-by-rule which is general practice for low-temperature applications. The rules laid down in the ASME Pressure Vessel Code Case N47 are explained as well as the procedure for inelastic stress calculations. (author)

  3. Configuration management after design basis reconstitution

    International Nuclear Information System (INIS)

    Purcell, J.J.; Livingston, B.R.

    1991-01-01

    Over the last few years, Fort Calhoun station (FCS) has implemented a number of programs to enhance plant operability and readiness. The design basis document (DBD) reconstitution project was the cornerstone of this effort. Vendor manual upgrade, operating procedures upgrade, plant equipment data-base verification, equipment labeling, and warehousing improvements were also implemented as part of this improvement program. With the completion of these programs, plant documentation was current to the baselines established by each program, and a configuration management program (CMP) was established to maintain this level of accuracy throughout the remaining life of FCS. Change control throughout the organization has been reviewed and upgraded to ensure that all changes are evaluated for impact to the design bases

  4. Acceptable risk as a basis for design

    International Nuclear Information System (INIS)

    Vrijling, J.K.; Hengel, W. van; Houben, R.J.

    1998-01-01

    Historically, human civilisations have striven to protect themselves against natural and man-made hazards. The degree of protection is a matter of political choice. Today this choice should be expressed in terms of risk and acceptable probability of failure to form the basis of the probabilistic design of the protection. It is additionally argued that the choice for a certain technology and the connected risk is made in a cost-benefit framework. The benefits and the costs including risk are weighed in the decision process. A set of rules for the evaluation of risk is proposed and tested in cases. The set of rules leads to technical advice in a question that has to be decided politically

  5. Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-11-01

    This report develops and applies a method for estimating strong earthquake ground motion. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Specifically considered are ground motions resulting from earthquakes with magnitudes from 5 to 8, fault distances from 0 to 500 km, and frequencies from 1 to 35 Hz. The two main objectives were: (1) to develop generic relations for estimating ground motion appropriate for site screening; and (2) to develop a guideline for conducting a thorough site investigation needed to define the seismic design basis. For the first objective, an engineering model was developed to predict the expected ground motion on rock sites, with an additional set of amplification factors to account for the response of the soil column over rock at soil sites. The results incorporate best estimates of ground motion as well as the randomness and uncertainty associated with those estimates. For the second objective, guidelines were developed for gathering geotechnical information at a site and using this information in calculating site response. As a part of this development, an extensive set of geotechnical and seismic investigations was conducted at three reference sites. Together, the engineering model and guidelines provide the means to select and assess the seismic suitability of a site

  6. 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)

  7. International Peer Reviews of Design Basis

    International Nuclear Information System (INIS)

    Hughes, Peter

    2013-01-01

    International peer reviews: Design and safety assessment review service: - Review of design requirements; - Review in support of licensing; - Review in support of severe accident management; - Review in support of modifications; - Review in relation to periodic safety, or life extension; - Reviews take place at any time in NPP lifecycle from concept, through design and operations

  8. Design basis for the NRC Operations Center

    Energy Technology Data Exchange (ETDEWEB)

    Lindell, M.K.; Wise, J.A.; Griffin, B.N.; Desrosiers, A.E.; Meitzler, W.D.

    1983-05-01

    This report documents the development of a design for a new NRC Operations Center (NRCOC). The project was conducted in two phases: organizational analysis and facility design. In order to control the amount of traffic, congestion and noise within the facility, it is recommended that information flow in the new NRCOC be accomplished by means of an electronic Status Information Management System. Functional requirements and a conceptual design for this system are described. An idealized architectural design and a detailed design program are presented that provide the appropriate amount of space for operations, equipment and circulation within team areas. The overall layout provides controlled access to the facility and, through the use of a zoning concept, provides each team within the NRCOC the appropriate balance of ready access and privacy determined from the organizational analyses conducted during the initial phase of the project.

  9. Design basis for the NRC Operations Center

    International Nuclear Information System (INIS)

    Lindell, M.K.; Wise, J.A.; Griffin, B.N.; Desrosiers, A.E.; Meitzler, W.D.

    1983-05-01

    This report documents the development of a design for a new NRC Operations Center (NRCOC). The project was conducted in two phases: organizational analysis and facility design. In order to control the amount of traffic, congestion and noise within the facility, it is recommended that information flow in the new NRCOC be accomplished by means of an electronic Status Information Management System. Functional requirements and a conceptual design for this system are described. An idealized architectural design and a detailed design program are presented that provide the appropriate amount of space for operations, equipment and circulation within team areas. The overall layout provides controlled access to the facility and, through the use of a zoning concept, provides each team within the NRCOC the appropriate balance of ready access and privacy determined from the organizational analyses conducted during the initial phase of the project

  10. A risk-informed framework for establishing a beyond design basis safety basis for external hazards

    Energy Technology Data Exchange (ETDEWEB)

    Amico, P. [Hughes Associates, Inc, Baltimore, MD (United States); Anoba, R. [Hughes Associates, Inc, Raleigh, NC (United States); Najafi, B. [Hughes Associates, Inc., Los Gatos, CA (United States)

    2014-07-01

    The events at Fukushima Daiichi taught us that meeting a deterministic design basis requirement for external hazards does not assure that the risk is low. As observed at the plant, the two primary reasons for this are failure cliffs above the design basis event and that combined hazard effects are not considered in design. Because the possible combinations of design basis exceedences and external hazard combinations are very large and complex, an approach focusing only on the most important ones is needed. For this reason, a risk informed approach is the most effective approach, which is discussed in this paper. (author)

  11. Data base pertinent to earthquake design basis

    International Nuclear Information System (INIS)

    Sharma, R.D.

    1988-01-01

    Mitigation of earthquake risk from impending strong earthquakes is possible provided the hazard can be assessed, and translated into appropriate design inputs. This requires defining the seismic risk problem, isolating the risk factors and quantifying risk in terms of physical parameters, which are suitable for application in design. Like all other geological phenomena, past earthquakes hold the key to the understanding of future ones. Quantificatio n of seismic risk at a site calls for investigating the earthquake aspects of the site region and building a data base. The scope of such investigations is il lustrated in Figure 1 and 2. A more detailed definition of the earthquake problem in engineering design is given elsewhere (Sharma, 1987). The present document discusses the earthquake data base, which is required to support a seismic risk evaluation programme in the context of the existing state of the art. (author). 8 tables, 10 figs., 54 refs

  12. Integral Monitored Retrievable Storage (MRS) Facility conceptual basis for design

    International Nuclear Information System (INIS)

    1985-10-01

    The purpose of the Conceptual Basis for Design is to provide a control document that establishes the basis for executing the conceptual design of the Integral Monitored Retrievable Storage (MRS) Facility. This conceptual design shall provide the basis for preparation of a proposal to Congress by the Department of Energy (DOE) for construction of one or more MRS Facilities for storage of spent nuclear fuel, high-level radioactive waste, and transuranic (TRU) waste. 4 figs., 25 tabs

  13. Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

    International Nuclear Information System (INIS)

    G. L. Sharp; R. T. McCracken

    2004-01-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety

  14. MOV motor and gearbox performance under design basis loads

    International Nuclear Information System (INIS)

    DeWall, K.G.; Watkins, J.C.

    1998-01-01

    This paper describes the results of valve testing sponsored by the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research and conducted at the Idaho National Engineering and Environmental Laboratory. The research objective was to evaluate the capabilities of specific actuator motor and gearbox assemblies under various design basis loading conditions. The testing was performed using the motor-operated valve load simulator, a test fixture that simulates the stem load profiles a valve actuator would experience when closing a valve against flow and pressure loadings. The authors tested five typical motors (four ac motors and one dc motor) with three gearbox assemblies at conditions a motor might experience in a power plant, including such off-normal conditions as operation at high temperature and reduced voltage. The authors also determined the efficiency of the actuator gearbox. The testing produced the following significant results: all five motors operated at or above their rated torque during tests at full voltage and ambient temperature; for all five motors (dc as well as ac), the actual torque loss due to voltage degradation was greater than the torque loss predicted using common methods; startup torques in locked rotor tests compared well with stall torques in dynamometer-type tests; the methods commonly used to predict torque losses due to elevated operating temperatures sometimes bounded the actual losses, but not in all cases; the greatest discrepancy involved the prediction for the dc motor; running efficiencies published by the manufacturer for actuator gearboxes were higher than the actual efficiencies determined from testing, in some instances, the published pullout efficiencies were also higher than the actual values; operation of the gearbox at elevated temperature did not affect the operating efficiency

  15. Upwind design basis (WP4 : Offshore foundations and support structures)

    NARCIS (Netherlands)

    Fischer, T.; De Vries, W.E.; Schmidt, B.

    2010-01-01

    The presented design basis gives a summarized overview of relevant design properties for a later offshore wind turbine design procedures within work package 4. The described offshore site is located in the Dutch North Sea and has a water depth of 21m. Therefore it will be chosen as shallow site

  16. SAFETY BASIS DESIGN DEVELOPMENT CHALLENGES IMECE2007-42747

    International Nuclear Information System (INIS)

    RYAN GW

    2007-01-01

    'Designing in Safety' is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation IOCFR830, Nuclear Safety Management. IOCFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses 'Safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE's Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, (1) flashing spray leaks and (2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents

  17. SAFETY BASIS DESIGN DEVELOPMENT CHALLENGES IMECE2007-42747

    Energy Technology Data Exchange (ETDEWEB)

    RYAN GW

    2007-09-24

    'Designing in Safety' is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation IOCFR830, Nuclear Safety Management. IOCFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses 'Safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE's Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, (1) flashing spray leaks and (2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents.

  18. Technical basis for the ITER-FEAT outline design

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-11-01

    This ITER EDA Documentation Series issue summarizes the results of the ITER Engineering Design Activities on the technical basis for the ITER-FEAT outline design. This issue also comprises some physical analysis activities as well as structure and goals of the Physics Expert Group activities.

  19. Technical basis for the ITER-FEAT outline design

    International Nuclear Information System (INIS)

    2000-01-01

    This ITER EDA Documentation Series issue summarizes the results of the ITER Engineering Design Activities on the technical basis for the ITER-FEAT outline design. This issue also comprises some physical analysis activities as well as structure and goals of the Physics Expert Group activities

  20. Solar Power Tower Design Basis Document, Revision 0

    Energy Technology Data Exchange (ETDEWEB)

    ZAVOICO,ALEXIS B.

    2001-07-01

    This report contains the design basis for a generic molten-salt solar power tower. A solar power tower uses a field of tracking mirrors (heliostats) that redirect sunlight on to a centrally located receiver mounted on top a tower, which absorbs the concentrated sunlight. Molten nitrate salt, pumped from a tank at ground level, absorbs the sunlight, heating it up to 565 C. The heated salt flows back to ground level into another tank where it is stored, then pumped through a steam generator to produce steam and make electricity. This report establishes a set of criteria upon which the next generation of solar power towers will be designed. The report contains detailed criteria for each of the major systems: Collector System, Receiver System, Thermal Storage System, Steam Generator System, Master Control System, and Electric Heat Tracing System. The Electric Power Generation System and Balance of Plant discussions are limited to interface requirements. This design basis builds on the extensive experience gained from the Solar Two project and includes potential design innovations that will improve reliability and lower technical risk. This design basis document is a living document and contains several areas that require trade-studies and design analysis to fully complete the design basis. Project- and site-specific conditions and requirements will also resolve open To Be Determined issues.

  1. Simulant Basis for the Standard High Solids Vessel Design

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-09-30

    The Waste Treatment and Immobilization Plant (WTP) is working to develop a Standard High Solids Vessel Design (SHSVD) process vessel. To support testing of this new design, WTP engineering staff requested that a Newtonian simulant and a non-Newtonian simulant be developed that would represent the Most Adverse Design Conditions (in development) with respect to mixing performance as specified by WTP. The majority of the simulant requirements are specified in 24590-PTF-RPT-PE-16-001, Rev. 0. The first step in this process is to develop the basis for these simulants. This document describes the basis for the properties of these two simulant types. The simulant recipes that meet this basis will be provided in a subsequent document.

  2. Information management needs for Fort Calhoun's design basis reconstitution project

    International Nuclear Information System (INIS)

    Beach, D.R.; Erickson, E.A.; Gambhir, S.K.; Parsons, R.D.

    1989-01-01

    While the need for information management is not new to the nuclear industry or Omaha Public Power District (OPPD), the interrelationship among design information, multiple systems, and design basis issues has necessitated the management of this information in new ways. The project team involved in the reconstitution of the design basis for OPPD's Fort Calhoun nuclear station has experienced the need for the developed effective methods for managing the vast amount of interrelated information associated with this effort. This management of information has been necessary to ensure that design basis documents (DBDs) adequately reflect the interrelated nature of component, system, and plant design; are complete and accurate; and are produced and maintained in a cost-effective manner. Fort Calhoun's aggressive design basis reconstitution project began in early 1987. The present scope of the project includes the production of 52 system and plant level DBDs; currently the project is ∼50% complete with DBDs in various stages of completion, from pilot DBDs through DBDs with approved formats, which have been issued for use. The experience in producing these documents has lead to a growing understanding of the special need for information management in each stage of the project. The development of the information tracking and management processes for the various stages of DBD development has proven to be cost-effective and gives a level of assurance that information has been included in the DBDs consistently and accurately

  3. Design basis document open-item resolution and reportability

    International Nuclear Information System (INIS)

    Gambhir, S.K.; Livingston, B.R.; Purcell, J.J.; Erickson, E.A.

    1989-01-01

    In the process of reconstituting the design bases for older nuclear power plants, information or references may not be available to fully define the design requirements or to document and verify the adequacy of the design. Also, information that is in conflict with other data is identified. The missing and conflicting information must be reconstituted in order to adequately document the design bases of the plant. For these operating facilities, the identification, tracking, and resolution of missing or conflicting information is very important when the reporting requirements stipulated by 10CFR21, 10CFR50.72, and 10CFR50.73 are considered. Additionally, controlled documentation (calculations, drawings, etc.) used to develop the design basis documents may contain conflicting data. In some cases, conflicts between the as-built design and licensing or design basis requirements established in specific commitments to the U.S. Nuclear Regulatory Commission may be identified. Furthermore, concerns regarding the adequacy of safety-related systems or components to perform their required function may be identified that would warrant prompt action by the licensee. The approach discussed in this paper was used by Omaha Public Power District for the ongoing design basis reconstitution effort at the Fort Calhoun nuclear plant

  4. Design basis programs and improvements in plant operation

    International Nuclear Information System (INIS)

    Metcalf, M.F.

    1991-01-01

    Public Service Electric and Gas (PSE and G) Company operates three commercial nuclear power plants in southern New Jersey. The three plants are of different designs and vintages (two pressurized water reactors licensed in 1976 and 1980 and one boiling water reactor licensed in 1986). As the industry recognized the need to develop design basis programs, PSE and G also realized the need after a voluntary 52-day shutdown of one unit because of electrical design basis problems. In its drive to be a premier electric utility, PSE and G has been aggressively active in developing design basis documents (DBDs) with supporting projects and refined uses to obtain the expected value and see the return on investment. Progress on Salem is nearly 75% complete, while Hope Creek is 20% complete. To data, PSE and G has experienced success in the use of DBDs in areas such as development of plant modifications, development of the reliability-centered maintenance program, procedure upgrades, improved document retrieval, resolution of regulatory issues, and training. The paper examines the design basis development process, supporting projects, and expected improvements in plant operations as a result of these efforts

  5. Design basis ground motion (Ss) required on new regulatory guide

    International Nuclear Information System (INIS)

    Kamae, Katsuhiro

    2013-01-01

    New regulatory guide is enforced on July 8. Here, it is introduced how the design basis ground motion (Ss) for seismic design of nuclear power reactor facilities was revised on the new guide. Ss is formulated as two types of earthquake ground motions, earthquake ground motions with site specific earthquake source and with no such specific source locations. The latter is going to be revised based on the recent observed near source ground motions. (author)

  6. Canister storage building design basis accident analysis documentation

    International Nuclear Information System (INIS)

    KOPELIC, S.D.

    1999-01-01

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  7. Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    PIEPHO, M.G.

    1999-10-20

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

  8. Canister storage building design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    KOPELIC, S.D.

    1999-02-25

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  9. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    International Nuclear Information System (INIS)

    CROWE, R.D.

    1999-01-01

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  10. Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

    International Nuclear Information System (INIS)

    PIEPHO, M.G.

    1999-01-01

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR

  11. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    International Nuclear Information System (INIS)

    CROWE, R.D.; PIEPHO, M.G.

    2000-01-01

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  12. Enhanced Design Alternative I: Low Temperature Design

    International Nuclear Information System (INIS)

    MacNeil, K.

    1999-01-01

    The purpose of this document is to evaluate Enhanced Design Alternative (EDA) 1, the low temperature repository design concept (CRWMS M and O 1999a). This technical document will provide supporting information for Site Recommendation (SR) and License Application (LA). Preparation of this evaluation will be in accordance with the technical document preparation plan (TDPP), (CRWMS M and O 1999b). EDA 1, one of five EDAs, was evolved from evaluation of a series of design features and alternatives developed during the first phase of the License Application Design Selection (LADS) process. Low, medium, and high temperature concepts were developed from the design features and alternatives prepared during Phase 1 of the LADS effort (CRWMS M and O 1999a). EDA 1 will first be evaluated against a single Screening Criterion, outlined in CRWMS M and O 1999a, which addresses post-closure performance of the repository. The performance of the repository is defined quantitatively as the peak radiological dose rate to an average individual of a critical group at a distance of 20 km from the repository site within 10,000 years. To satisfy this criterion the peak dose rate must not exceed the anticipated regulatory level of 25 mrem/yr within 10,000 years. If the EDA meets the screening criterion, the EDA will be further evaluated against the LADS Phase 2 Evaluation Criteria contained in CRWMS M and O 1999a

  13. The RCC-MR design code for LMFBR components. A useful basis for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1986-01-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials, temperature service level, loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain. (author)

  14. Basis for NGNP Reactor Design Down-Selection

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2010-08-01

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  15. Technical Details on Beyond Design Basis Event Pilot Evaluations

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2013-01-01

    The primary focus of the BDBE pilot project was the review of BDBE analysis and mitigation features at four DOE nuclear facilities representing a range of DOE sites, nuclear facility types/activities, and responsible program offices. The pilots looked at (1) how beyond design basis accidents were evaluated and documented in the facility Documented Safety Analysis, (2) potential BDBE vulnerabilities and margins to failure of facility safety features as obtained from general area and specific system walkdowns and design documents reviews, and (3) preparations made in facility and site emergency management programs to respond to severe accidents. It also evaluated whether draft BDBE guidance on safety analysis and emergency management could be used to improve the analysis of and preparations for mitigating severe and beyond design basis accidents. The details of these activities are organized in this report as described below.

  16. Design-Load Basis for LANL Structures, Systems, and Components

    Energy Technology Data Exchange (ETDEWEB)

    I. Cuesta

    2004-09-01

    This document supports the recommendations in the Los Alamos National Laboratory (LANL) Engineering Standard Manual (ESM), Chapter 5--Structural providing the basis for the loads, analysis procedures, and codes to be used in the ESM. It also provides the justification for eliminating the loads to be considered in design, and evidence that the design basis loads are appropriate and consistent with the graded approach required by the Department of Energy (DOE) Code of Federal Regulation Nuclear Safety Management, 10, Part 830. This document focuses on (1) the primary and secondary natural phenomena hazards listed in DOE-G-420.1-2, Appendix C, (2) additional loads not related to natural phenomena hazards, and (3) the design loads on structures during construction.

  17. Advances in the physics basis for the European DEMO design

    Science.gov (United States)

    Wenninger, R.; Arbeiter, F.; Aubert, J.; Aho-Mantila, L.; Albanese, R.; Ambrosino, R.; Angioni, C.; Artaud, J.-F.; Bernert, M.; Fable, E.; Fasoli, A.; Federici, G.; Garcia, J.; Giruzzi, G.; Jenko, F.; Maget, P.; Mattei, M.; Maviglia, F.; Poli, E.; Ramogida, G.; Reux, C.; Schneider, M.; Sieglin, B.; Villone, F.; Wischmeier, M.; Zohm, H.

    2015-06-01

    In the European fusion roadmap, ITER is followed by a demonstration fusion power reactor (DEMO), for which a conceptual design is under development. This paper reports the first results of a coherent effort to develop the relevant physics knowledge for that (DEMO Physics Basis), carried out by European experts. The program currently includes investigations in the areas of scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.

  18. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    1999-09-09

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  19. Design basis tropical cyclone for nuclear power plants

    International Nuclear Information System (INIS)

    1984-01-01

    The general characteristics of tropical cyclones are discussed in this Safety Guide, with particular emphasis on their pressure and wind structures in the light of available data. General methods are given for the evaluation of the relevant parameters of a Probable Maximum Tropical Cyclone (PMTC), which can be used as the Design Basis Tropical Cyclone (DBTC); these parameters then serve as inputs for the derivation of a design basis surge and a design basis wind. A possible method is also given for the evaluation of the PMTC pressure and wind field based on an approach valid primarily for a particular region. This method depends on the results of a theoretical study on the tropical cyclone structure and makes use of a large amount of data, including aircraft reconnaissance observations for 170 most intense tropical cyclones near the coast of Japan, Taiwan and the Philippines for the period 1960-1974, as well as detailed analyses of all the extreme storms along the Gulf of Mexico and the east coast of the USA during 1900-1978, for the determination of the necessary parameters

  20. High temperature fusion reactor design

    International Nuclear Information System (INIS)

    Harkness, S.D.; dePaz, J.F.; Gohar, M.Y.; Stevens, H.C.

    1979-01-01

    Fusion energy may have unique advantages over other systems as a source for high temperature process heat. A conceptual design of a blanket for a 7 m tokamak reactor has been developed that is capable of producing 1100 0 C process heat at a pressure of approximately 10 atmospheres. The design is based on the use of a falling bed of MgO spheres as the high temperature heat transfer system. By preheating the spheres with energy taken from the low temperature tritium breeding part of the blanket, 1086 MW of energy can be generated at 1100 0 C from a system that produces 3000 MW of total energy while sustaining a tritium breeding ratio of 1.07. The tritium breeding is accomplished using Li 2 O modules both in front of (6 cm thick) and behind (50 cm thick) the high temperature ducts. Steam is used as the first wall and front tritium breeding module coolant while helium is used in the rear tritium breeding region. The system produces 600 MW of net electricity for use on the grid

  1. Design basis event consequence analyses for the Yucca Mountain project

    International Nuclear Information System (INIS)

    Orvis, D.D.; Haas, M.N.; Martin, J.H.

    1997-01-01

    Design basis event (DBE) definition and analysis is an ongoing and integrated activity among the design and analysis groups of the Yucca Mountain Project (YMP). DBE's are those that potentially lead to breach of the waste package and waste form (e.g., spent fuel rods) with consequent release of radionuclides to the environment. A Preliminary Hazards Analysis (PHA) provided a systematic screening of external and internal events that were candidate DBE's that will be subjected to analyses for radiological consequences. As preparation, pilot consequence analyses for the repository subsurface and surface facilities have been performed to define the methodology, data requirements, and applicable regulatory limits

  2. Emergency procedures beyond design basis ''Feed and Bleed''

    International Nuclear Information System (INIS)

    Dominguez Bautista, M.T.; Campuzano Pena, F.

    1994-01-01

    The incorporation of Beyond-Design-Basis Emergency Procedures, also called the Emergency Manual or Severe Accident Manual, has been an important step forward in nuclear power plant safety. These procedures cover situations in which the deterministic criteria used in plant design have been contravened. In such situations new accident scenarios, unforeseen system actions or a combination of both, need to be considered. Establishing these procedures is actually the last in a sequence of activities the sequence includes definition of scenarios, study of their phenomena, analysis of optional system actions, verification of their effectiveness and finally, implementation of the procedure. The systematization of these new strategies is supported by the results of the probabilistic analyses which serve in this case to pinpoint the objectives of these strategies. This paper describes the application of this methodology in the definition of a procedure for heat sink recovery on the secondary side (feed and bleed) if this has been totally or partially lost in a beyond-design-basis event. (Author)

  3. Spent Nuclear Fuel (SNF) Project Design Basis Capacity Study

    International Nuclear Information System (INIS)

    CLEVELAND, K.J.

    2000-01-01

    This study of the design basis capacity of process systems was prepared by Fluor Federal Services for the Spent Nuclear Fuel Project. The evaluation uses a summary level model of major process sub-systems to determine the impact of sub-system interactions on the overall time to complete fuel removal operations. The process system model configuration and time cycle estimates developed in the original version of this report have been updated as operating scenario assumptions evolve. The initial document released in Fiscal Year (FY) 1996 varied the number of parallel systems and transport systems over a wide range, estimating a conservative design basis for completing fuel processing in a two year time period. Configurations modeling planned operations were updated in FY 1998 and FY 1999. The FY 1998 Base Case continued to indicate that fuel removal activities at the basins could be completed in slightly over 2 years. Evaluations completed in FY 1999 were based on schedule modifications that delayed the start of KE Basin fuel removal, with respect to the start of KW Basin fuel removal activities, by 12 months. This delay resulted in extending the time to complete all fuel removal activities by 12 months. However, the results indicated that the number of Cold Vacuum Drying (CVD) stations could be reduced from four to three without impacting the projected time to complete fuel removal activities. This update of the design basis capacity evaluation, performed for FY 2000, evaluates a fuel removal scenario that delays the start of KE Basin activities such that staffing peaks are minimized. The number of CVD stations included in all cases for the FY 2000 evaluation is reduced from three to two, since the scenario schedule results in minimal time periods of simultaneous fuel removal from both basins. The FY 2000 evaluation also considers removal of Shippingport fuel from T Plant storage and transfer to the Canister Storage Building for storage

  4. Archaeological data as a basis for repository marker design

    International Nuclear Information System (INIS)

    Kaplan, M.F.

    1982-10-01

    This report concerns the development of a marking system for a nuclear waste repository which is very likely to survive for 10,000 years. In order to provide a background on the subject, and for the preliminary design presented in this report, a discussion is presented about the issues involved in human interference with the repository system and the communication of information. A separate chapter summarizes six ancient man-made monuments including: materials, effects of associated textual information on our understanding of the monument, and other features of the ancient monument relevant to marking a repository site. The information presented in the two chapters is used to provide the basis and rationale for a preliminary marker system design presented in a final chapter. 86 refs., 22 figs., 1 tab

  5. Reactor safety under design basis flood condition for inland sites

    International Nuclear Information System (INIS)

    Hajela, S.; Bajaj, S.S.; Samota, A.; Verma, U.S.P.; Warudkar, A.S.

    2002-01-01

    Full text: In June 1994, there was an incident of flooding at Kakrapar Atomic Power Station (KAPS) due to combination of heavy rains and mechanical failure in the operation of gates at the adjoining weir. An indepth review of the incident was carried out and a number of flood protection measures were recommended and were implemented at site. As part of this review, a safety analysis was also done to demonstrate reactor safety with a series of failures considered in the flood protection features. For each inland NPP site, as part of design, different flood scenarios are analysed to arrive at design basis flood (DBF) level. This level is estimated based on worst combination of heavy local precipitation, flooding in river, failure of upstream/downstream water control structures

  6. Archaeological data as a basis for repository marker design

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, M.F.

    1982-10-01

    This report concerns the development of a marking system for a nuclear waste repository which is very likely to survive for 10,000 years. In order to provide a background on the subject, and for the preliminary design presented in this report, a discussion is presented about the issues involved in human interference with the repository system and the communication of information. A separate chapter summarizes six ancient man-made monuments including: materials, effects of associated textual information on our understanding of the monument, and other features of the ancient monument relevant to marking a repository site. The information presented in the two chapters is used to provide the basis and rationale for a preliminary marker system design presented in a final chapter. 86 refs., 22 figs., 1 tab.

  7. Cold Vacuum Drying facility design basis accident analysis documentation

    International Nuclear Information System (INIS)

    CROWE, R.D.

    2000-01-01

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls

  8. Design basis for resistance to shock and vibration

    International Nuclear Information System (INIS)

    Glass, R.E.; Gwinn, K.W.

    1989-01-01

    Sandia National Laboratories, in conjunction with its participation in the American National Standards Institute (ANSI) writing groups, has undertaken to provide an experimental and analytical basis for the design of components of radioactive materials packages to resist normal transport shock and vibration loads. Previous efforts have resulted in an overly conservative shock spectra description of the loads in the tie-downs and cask attachment points anticipated during normal shipment. The present effort is aimed at predicting the actual loads so that the design basis can be accurately determined. This goal is being accomplished with road simulator and over-the-road tests and the development of an analytical model. This model is used to parametrically evaluate and envelop the transportation systems' responses. The parameters to be varied include damping, stiffness, geometry, and cargo mass. The over-the-road tests provide operational data that are used to validate the selection of environments for the road simulator tests. The road simulator tests provide verification for the model. This verification is accomplished since the road simulator tests provide not only the system response which can be measured in over-the-road tests but also the system input. Finally, when the model has been verified, it can be used to vary parameters to envelop a wide range of normal transport conditions

  9. Cold Vacuum Drying facility design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    2000-08-08

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

  10. Design basis for resistance to shock and vibration

    International Nuclear Information System (INIS)

    Glass, R.E.; Gwinn, K.W.

    1989-01-01

    Sandia National Laboratories, in conjunction with its participation in the American National Standards Institute (ANSI) writing groups, has undertaken to provide an experimental and analytical basis for the design of components of radioactive materials packages to resist normal transport shock and vibration loads. Previous efforts have resulted in an overly conservative shock spectra description of the loads in the tie-downs and cask attachment points anticipated during normal shipment. The present effort is aimed at predicting the actual loads so that the design basis can be accurately determined. This goal is being accomplished with road simulator and over-the-road tests and the development of an analytical model. This model is used to parametrically evaluate and envelop the transportation systems responses. The parameters to be varied include damping, stiffness, geometry, and cargo mass. The over-the-road tests provide operational data that are used to validate the selection of environments for the road simulator tests. The road simulator tests provide verification for the model. This verification is accomplished since the road simulator tests provide not only the system response which can be measured in over-the-road tests but also the system input. Finally, when the model has been verified, it can be used to vary parameters to envelope a wide range of normal transport conditions

  11. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  12. Transient and accident analyses topical design basis documents

    International Nuclear Information System (INIS)

    Chi, Larry; Eckert, Eugene; Grim, Brit

    2004-01-01

    The designers and operators of nuclear power plants have extensively documented system functions, licensing performance, and operating procedures for all conditions. This paper presents a complementary, systematic approach for the documentation of all requirements that are based on the analysis of operational transients, abnormal transients, accidents, and other events which are included in the design and licensing basis for the plant. Up to now, application of the approach has focused on required mitigation actions (automatic or manual). All mitigation actions are directly identified with all applicable reactor events, as well as the plant-unique systems that work together to perform each function. The approach is also applicable to all operational functions. The approach makes extensive use of data base methods, thereby providing effective ways to interrogate the information for the varied users of this information. Examples of use include: evaluations of system design changes and equipment modifications, safety evaluations of any plant change (e.g., USNRC 10CFR50.59 review), plant operations (e.g., manual actions during unplanned events), system interactions, classification of safety-related equipment, environmental qualification of equipment, and mitigation requirements for different reactor operating states. This approach has been applied in customized ways to several boiling water reactor (BWR) units, based on the desires and needs of the specific utility. (author)

  13. Data Requirements and the Basis for Designing Health Information Kiosks.

    Science.gov (United States)

    Afzali, Mina; Ahmadi, Maryam; Mahmoudvand, Zahra

    2017-09-01

    Health kiosks are an innovative and cost-effective solution that organizations can easily implement to help educate people. To determine the data requirements and basis for designing health information kiosks as a new technology to maintain the health of society. By reviewing the literature, a list of information requirements was provided in 4 sections (demographic information, general information, diagnostic information and medical history), and questions related to the objectives, data elements, stakeholders, requirements, infrastructures and the applications of health information kiosks were provided. In order to determine the content validity of the designed set, the opinions of 2 physicians and 2 specialists in medical informatics were obtained. The test-retest method was used to measure its reliability. Data were analyzed using SPSS software. In the proposed model for Iran, 170 data elements in 6 sections were presented for experts' opinion, which ultimately, on 106 elements, a collective agreement was reached. To provide a model of health information kiosk, creating a standard data set is a critical point. According to a survey conducted on the various literature review studies related to the health information kiosk, the most important components of a health information kiosk include six categories; information needs, data elements, applications, stakeholders, requirements and infrastructure of health information kiosks that need to be considered when designing a health information kiosk.

  14. DART - for design basis justification and safety related information management

    International Nuclear Information System (INIS)

    Billington, A.; Blondiaux, P.; Boucau, J.; Cantineau, B.; Doumont, C.; Mared, A.

    2000-01-01

    DART is the acronym for Design Analysis Re-engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

  15. DART - for design basis justification and safety related information management

    International Nuclear Information System (INIS)

    Billington, A.; Blondiaux, B.; Boucau, J.; Cantineau, B.; Mared, A.

    2001-01-01

    DART is the acronym for Design Analysis Re-Engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can then be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

  16. Study on risk factors of PWR accidents beyond design basis

    International Nuclear Information System (INIS)

    Ahn, Seung Hoon; Nah, W. J.; Bang, Y. S.; Oh, D. Y.; Oh, S. H.

    2005-01-01

    Development of the regulatory guidelines for Beyond Design Basis Accidents (BDBA) with high risk requires a detailed investigation of major factors contributing to the event risk. In this study, each event was classified by the level of risk, based on the probabilistic safety assessment results, so that BDBA with high risk could be selected, with consideration of foreign and domestic regulations, and operating experiences. The regulatory requirements and technical backgrounds for the selected accidents were investigated, and effective regulatory approaches for risk reduction of the accidents. The following conclusions were drawn from this study: - Selected high risk BDBA is station blackout, anticipated without scram, total loss of feedwater. - Major contributors to the risk of selected events were investigated, and appropriate assessment of them was recommended for development of the regulatory guidelines

  17. Technical basis for the ITER-FEAT outline design. Progress in resolving open design issues from the outline design report

    International Nuclear Information System (INIS)

    2000-01-01

    In this publication the technical basis for the ITER-FEAT outline design is presented. It comprises the Plant Design Specifications, the Safety Principles and Environmental Criteria, the Site Requirements and Site Design Assumptions. The outline of the key features of the ITER-FEAT design includes main physical parameters and assessment, design overview and preliminary safety assessment, cost and schedule

  18. The concept of risk in the design basis threat

    International Nuclear Information System (INIS)

    Reynolds, J.M.

    2001-01-01

    Full text: Mathematically defined, risk is a product of one or more probability factors and one or more consequences. Actuarial analysis of risk requires the creation of a numeric algorithm that reflects the interaction of different probability factors, where probability data usually draws on direct measurements of incidence. For physical protection purposes, the algorithms take the general form: Risk = Probability of successful attack x Consequence where the overall probability of a successful attack will be determined by the product of, amongst other things, the probability of there being sufficient intent, the probability of there being available hostile resources, the probability of deterrence, and the probability that a hostile act will be detected and prevented. Deliberate, malevolent acts against nuclear facilities are rare. In so far as it is possible to make an actuarial type of judgement, the probability of malevolent activity against a nuclear facility is almost zero. This creates a problem for a numerical assessment of risk for nuclear facilities where the value (consequence) term could be almost infinite. As can be seen from the general equation above, a numerical algorithm of risk of malevolent activity affecting nuclear facilities could only yield a zero or infinite result. In such circumstances, intelligence-based threat assessments are sometimes thought of as a substitute for historic data in the determination of probability. However, if the paucity of historic data reflects the actual threat - which by and large it should - no amount of intelligence is likely to yield a substantially different conclusion. This mathematical approach to analysing risk appears to lead us either to no risk and no protection or to an infinite risk demanding every conceivable protective measure. The Design Basis Threat (DBT) approach offers a way out of the dilemma. Firstly, it allows us to eliminate from further consideration all zero or near zero probabilities

  19. Design basis flood for nuclear power plants on river sites

    International Nuclear Information System (INIS)

    1983-01-01

    The Guide presents techniques for determining the design basis flood (DBF) to be used for siting nuclear power plants at or near non-tidal reaches of rivers and for protecting nuclear power plants against floods. Since flooding of a nuclear power plant can have repercussions on safety, the DBF is always chosen to have a very low probability of exceedance per annum. The DBF may result from one or more of the following causes: (1) Precipitation, snowmelt; (2) Failure of water control structures, either from seismic or hydrological causes or from faulty operation of these structures; (3) Channel obstruction such as landslide, ice effects, log or debris jams, and effects of vulcanism. Normally the DBF is not less than any recorded or historical flood occurrence. For flood evaluation two types of methods are discussed in this Guide: probabilistic and deterministic. Simple probabilistic methods to determine floods of such low exceedance probability have a great degree of uncertainty and are presented for use only during the site survey. However, the more sophisticated probabilistic methods, the so-called stochastic methods, may give an acceptable result, as outlined in this Guide. The preferred method of evaluating the component of the DBF due to precipitation, as described in this Guide, is the deterministic one, based on the concept of a limit to the probable maximum precipitation (PMP) and on the unit hydrograph technique. Dam failures may generate a flood substantially more severe than that due to precipitation. The methodology for evaluating these types of floods is therefore presented in this Guide. Making allowance for the possible simultaneous occurrence of two or more important flood-producing events is also discussed here. The Guide does not deal with floods caused by sabotage

  20. Design of a dynamic compensated temperature sensor

    International Nuclear Information System (INIS)

    Yan, Wu; Katz, E.M.; Kerlin, T.W.

    1991-01-01

    One important function of a temperature sensor in a nuclear power plant is to track changing process temperatures, but the sensor output lags the changing temperature. This lag may have a large influence when the sensor is used in control or safety systems. Therefore, it is advantageous to develop methods that increase the sensor response speed. The goal of this project is to develop a fast-responding temperature sensor, the dynamic compensated temperature sensor (DCTS), based on signal dynamic compensation technology. To verify the theoretical basis of the DCTS and incorporate the DCTS into a real temperature measurement process, several experiments have been performed. The DCTS is a simple approach that can decrease the temperature sensor's response time, and it can provide faster temperature signals to the nuclear power plant safety system

  1. Study of elevated temperature design standard against thermal loads

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Asayama, Tai; Morishita, Masaki

    2001-01-01

    Elevated temperature components must be designed against both pressure and thermal loads. In the case of sodium circuits of fast breeder reactors, a restriction from the pressure load becomes small because of the high boiling point of sodium. Design approaches for thermal loads (displacement-controlled) are compared with those against pressure loads (load-controlled). Considering differences between those two approaches, a concept of the elevated temperature design standard that takes the nature of thermal loads fully into account is proposed. This concept is a basis of load evaluation techniques and an inelastic analysis guide, that are being developed. Finally, problems and plans to realize the above concept are discussed. (author)

  2. Reconstruction of Daily Sea Surface Temperature Based on Radial Basis Function Networks

    Directory of Open Access Journals (Sweden)

    Zhihong Liao

    2017-11-01

    Full Text Available A radial basis function network (RBFN method is proposed to reconstruct daily Sea surface temperatures (SSTs with limited SST samples. For the purpose of evaluating the SSTs using this method, non-biased SST samples in the Pacific Ocean (10°N–30°N, 115°E–135°E are selected when the tropical storm Hagibis arrived in June 2014, and these SST samples are obtained from the Reynolds optimum interpolation (OI v2 daily 0.25° SST (OISST products according to the distribution of AVHRR L2p SST and in-situ SST data. Furthermore, an improved nearest neighbor cluster (INNC algorithm is designed to search for the optimal hidden knots for RBFNs from both the SST samples and the background fields. Then, the reconstructed SSTs from the RBFN method are compared with the results from the OI method. The statistical results show that the RBFN method has a better performance of reconstructing SST than the OI method in the study, and that the average RMSE is 0.48 °C for the RBFN method, which is quite smaller than the value of 0.69 °C for the OI method. Additionally, the RBFN methods with different basis functions and clustering algorithms are tested, and we discover that the INNC algorithm with multi-quadric function is quite suitable for the RBFN method to reconstruct SSTs when the SST samples are sparsely distributed.

  3. The earthquake problem in engineering design: generating earthquake design basis information

    International Nuclear Information System (INIS)

    Sharma, R.D.

    1987-01-01

    Designing earthquake resistant structures requires certain design inputs specific to the seismotectonic status of the region, in which a critical facility is to be located. Generating these inputs requires collection of earthquake related information using present day techniques in seismology and geology, and processing the collected information to integrate it to arrive at a consolidated picture of the seismotectonics of the region. The earthquake problem in engineering design has been outlined in the context of a seismic design of nuclear power plants vis a vis current state of the art techniques. The extent to which the accepted procedures of assessing seismic risk in the region and generating the design inputs have been adherred to determine to a great extent the safety of the structures against future earthquakes. The document is a step towards developing an aproach for generating these inputs, which form the earthquake design basis. (author)

  4. Reduced design load basis for ultimate blade loads estimation in multidisciplinary design optimization frameworks

    DEFF Research Database (Denmark)

    Pavese, Christian; Tibaldi, Carlo; Larsen, Torben J.

    2016-01-01

    The aim is to provide a fast and reliable approach to estimate ultimate blade loads for a multidisciplinary design optimization (MDO) framework. For blade design purposes, the standards require a large amount of computationally expensive simulations, which cannot be efficiently run each cost...... function evaluation of an MDO process. This work describes a method that allows integrating the calculation of the blade load envelopes inside an MDO loop. Ultimate blade load envelopes are calculated for a baseline design and a design obtained after an iteration of an MDO. These envelopes are computed...... for a full standard design load basis (DLB) and a deterministic reduced DLB. Ultimate loads extracted from the two DLBs with the two blade designs each are compared and analyzed. Although the reduced DLB supplies ultimate loads of different magnitude, the shape of the estimated envelopes are similar...

  5. Licensing topical report: application of probabilistic risk assessment in the selection of design basis accidents

    International Nuclear Information System (INIS)

    Houghton, W.J.

    1980-06-01

    A probabilistic risk assessment (PRA) approach is proposed to be used to scrutinize selection of accident sequences. A technique is described in this Licensing Topical Report to identify candidates for Design Basis Accidents (DBAs) utilizing the risk assessment results. As a part of this technique, it is proposed that events with frequencies below a specified limit would not be candidates. The use of the methodology described is supplementary to the traditional, deterministic approach and may result, in some cases, in the selection of multiple failure sequences as DBAs; it may also provide a basis for not considering some traditionally postulated events as being DBAs. A process is then described for selecting a list of DBAs based on the candidates from PRA as supplementary to knowledge and judgments from past licensing practice. These DBAs would be the events considered in Chapter 15 of Safety Analysis Reports of high-temperature gas-cooled reactors

  6. Development of Mitigation Strategy for Beyond Design Basis External Events for NRC Design Certification

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Hak; Lee, Jae Jong; Kim, Myung Ki [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, how to develop FLEX strategy for beyond-design-basis external events for U. S. NRC design certification is examined. The development method of FLEX strategy for U. S. NRC design certification is examined. The applicants should make unit-specific FLEX strategy and establish the minimum coping capabilities consistent with unit-specific evaluation of the potential impacts and responses to BDBEEs. NEI 12-06 outlines the process to define and deploy the diverse and flexible mitigation strategies(FLEX strategy) that will increase defense-in-depth for beyond-design-basis scenarios to address the extended loss of alternating current (ac) power (ELAP) and loss of normal access to the ultimate heat sink (LUHS) occurring simultaneously at all units on a site. The order (EA-12-049) is issued to all reactor licensees, including holders of active, Construction Permit (CP) holders, and Combined License (COL) holders. Applicants for the new reactor design certification should prepare and submit FLEX strategy for NRC staff's review. Site-specific data related with the new reactor can't be determined during the new reactor design certification applications so that the unit-specific FLEX strategy should be developed.

  7. Development of Mitigation Strategy for Beyond Design Basis External Events for NRC Design Certification

    International Nuclear Information System (INIS)

    Kim, Dong Hak; Lee, Jae Jong; Kim, Myung Ki

    2013-01-01

    In this study, how to develop FLEX strategy for beyond-design-basis external events for U. S. NRC design certification is examined. The development method of FLEX strategy for U. S. NRC design certification is examined. The applicants should make unit-specific FLEX strategy and establish the minimum coping capabilities consistent with unit-specific evaluation of the potential impacts and responses to BDBEEs. NEI 12-06 outlines the process to define and deploy the diverse and flexible mitigation strategies(FLEX strategy) that will increase defense-in-depth for beyond-design-basis scenarios to address the extended loss of alternating current (ac) power (ELAP) and loss of normal access to the ultimate heat sink (LUHS) occurring simultaneously at all units on a site. The order (EA-12-049) is issued to all reactor licensees, including holders of active, Construction Permit (CP) holders, and Combined License (COL) holders. Applicants for the new reactor design certification should prepare and submit FLEX strategy for NRC staff's review. Site-specific data related with the new reactor can't be determined during the new reactor design certification applications so that the unit-specific FLEX strategy should be developed

  8. Design Basis for Fibre Reinforced Concrete (FRC) Pavements

    DEFF Research Database (Denmark)

    Bendixen, Søren; Stang, Henrik

    1996-01-01

    -crack opening relationship can beused to descibe the properties of fibre reinforced concrete (FRC) intension and how the stress-crack opening relationship can beapplied in a simple design scheme for pavements. The projectincludes development of design tools, experiments to determine thestress-crack opening...

  9. Key issues on safety design basis selection and safety assessment

    International Nuclear Information System (INIS)

    An, S.; Togo, Y.

    1976-01-01

    In current fast reactor design in Japan, four design accident conditions and four design seismic conditions are adopted as the design base classifications. These are classified by the considerations on both likelihood of occurrence and the severeness of the consequences. There are several major problem areas in safety design consideration such as core accident problems which include fuel sodium interaction, fuel failure propagation and residual decay heat removal, and decay heat removal systems problems which is more or less the problem of selection of appropriate system and of assurance of high reliability of the system. In view of licensing, two kinds of accidents are postulated in evaluating the adequacy of a reactor site. The one is the ''major accident'' which is the accident to give most severe radiation hazard to the public from technical point of view. The other is the ''hypothetical accident'', induced public accident of which is severer than that of major accident. While the concept of the former is rather unique to Japanese licensing, the latter is almost equivalent to design base hypothetical accident of the US practice. In this paper, design bases selections, key safety issues and some of the licensing considerations in Japan are described

  10. CIF---Design basis for an integrated incineration facility

    International Nuclear Information System (INIS)

    Bennett, G.F.

    1991-01-01

    This paper discusses the evolution of chosen technologies that occurred during the design process of the US Department of Energy (DOE) incineration system designated the Consolidated Incineration Facility (CIF) as the Savannah River Plant, Aiken, South Carolina. The Plant is operated for DOE by the Westinghouse Savannah River Company. The purpose of the incineration system is to treat low level radioactive and/or hazardous liquid and solid wastes by combustion. The objective for the facility is to thermally destroy toxic constituents and volume reduce waste material. Design criteria requires operation be controlled within the limits of RCRA's permit envelope

  11. Accident beyond the design basis management with the coolant loss at the NPP with WWER

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Klyuchnikov, A.A.; Kolykhanov, V.N.

    2010-01-01

    The analysis of status and experience of development on modelling and accident beyond the design basis management, including the severe accidents, at the nuclear power plants is carried out. The methodical providing of manuals on the accident beyond the design basis management with the coolant loss on the basis of simulated critical system configurations providing the necessary safety function performance on reactor unit is proposed. The project of symptom-oriented manuals on accident beyond the design basis management with the coolant loss on the serial power unit with WWER-1000 on the basis of developed methodical providing and well known results of deepened safety analysis is presented.

  12. Designing an accurate system for temperature measurements

    Directory of Open Access Journals (Sweden)

    Kochan Orest

    2017-01-01

    Full Text Available The method of compensation of changes in temperature field along the legs of inhomogeneous thermocouple, which measures a temperature of an object, is considered in this paper. This compensation is achieved by stabilization of the temperature field along the thermocouple. Such stabilization does not allow the error due to acquired thermoelectric inhomogeneity to manifest itself. There is also proposed the design of the furnace to stabilize temperature field along the legs of the thermocouple which measures the temperature of an object. This furnace is not integrated with the thermocouple mentioned above, therefore it is possible to replace this thermocouple with a new one when it get its legs considerably inhomogeneous.. There is designed the two loop measuring system with the ability of error correction which can use simultaneously a usual thermocouple as well as a thermocouple with controlled profile of temperature field. The latter can be used as a reference sensor for the former.

  13. Design basis for the operational modelling of the atmospheric dispersion

    International Nuclear Information System (INIS)

    Doury, A.

    1987-10-01

    Based on the latest practices at the Institut de Protection et de Surete Nucleaire of the Commissariat a l'Energie Atomique (CEA), we shall first present the basis elements used for a simple and adequate modelling method for assessing hypothetical atmospheric pollution from transient or continuous discharge with any given kinetics under various weather conditions which are not necessarily stationary or uniform, which are likely to occur even with little or no wind. Discharges shall be considered as sequences of instantaneous successive puffs. The parameters deduced experimentally or from observations are functions of the transfer time and cover all time and space scales. The restrictions of use are indicated, especially concerning heavy gases. Finally, simple formulas are proposed for concentrations and depositions so as to be able to make a rapid estimation of the orders of magnitude with almost no computation [fr

  14. Design basis for the operational modelling of the atmospheric dispersion

    International Nuclear Information System (INIS)

    Doury, A.

    1987-11-01

    Based on the latest practices at the Institut de Protection et de Surete Nucleaire of the Commissariat a l'Energie Atomique (CEA), we shall first present the basis elements used for a simple and adequate modelling method for assessing hypothetical atmospheric pollution from transient or continuous discharge with any given kinetics under various weather conditions which are not necessarily stationary or uniform, which are likely to occur even with little or no wind. Discharges shall be considered as sequences of instantaneous successive puffs. The parameters deduced experimentally or from observations are functions of the transfer time and cover all time and space scales. The restrictions of use are indicated, especially concerning heavy gases. Finally, simple formulas are proposed for concentrations and depositions so as to be able to make a rapid estimation of the orders of magnitude with almost no computation [fr

  15. Design basis reconstitution and configuration management of nuclear power plants

    International Nuclear Information System (INIS)

    Smith, P.R.

    1989-01-01

    The major design requirements of nuclear power plant components, systems, and structures are found in the plant's licensing commitments documented in the Final Safety Analysis Report and in the technical specification commitments of the plant. These specifications consider the original design and its degradation by in-service use. Before a nuclear power plant begins operation, the plant systems, structures, and organizational elements are functionally arranged to operate in a particular way. This functional arrangement is specified by the plant's design requirements and is called its configuration. The paper discusses configuration management and information management for configuration management. The management of large amounts of information and the various information systems associated with nuclear generating facilities is an ever-growing challenge for utilities. Plant operations involve a complex interrelation among data elements, especially in relation to design modifications and operational changes. Consequently, the operation of these data systems is interrelated and, as a result, redundant data items may exist. Thus, in view of the need to control and manage the plant configuration baseline, managers are striving to streamline their information management programs, which usually involves the integration of data-base systems

  16. Simulant Basis for the Standard High Solids Vessel Design

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-01

    This document provides the requirements for a test simulant suitable for demonstrating the mixing requirements for the Single High Solids Vessel Design (SHSVD). This simulant has not been evaluated for other purposes such as gas retention and release or erosion. The objective of this work is to provide an underpinning for the simulant properties based on actual waste characterization.

  17. Cognitive Process as a Basis for Intelligent Retrieval Systems Design.

    Science.gov (United States)

    Chen, Hsinchun; Dhar, Vasant

    1991-01-01

    Two studies of the cognitive processes involved in online document-based information retrieval were conducted. These studies led to the development of five computational models of online document retrieval which were incorporated into the design of an "intelligent" document-based retrieval system. Both the system and the broader implications of…

  18. Concepts on high temperature design analysis for SNR 300

    International Nuclear Information System (INIS)

    Bieniussa, K.; Zolti, E.

    1976-01-01

    The paper briefly describes the evolution, the present situation and the next activities on the design of high temperature components of the DEBENELUX prototype fast breeder reactor SNR-300 with particular regard to the design criteria. Elastic structural analyses are performed for the basic design of the components and are supplied by the manufacturer. In agreement with the Safety Experts simplified and/or detailed inelastic analyses of the critical areas are supplied by the prime contractor of the plant. The elastic computations are evaluated on the basis of a set of design rules derived from ASME Code Case Interpretation 1331-4 but with more conservative limits, and the inelastic ones on the basis of the ASME Code Case Interpretation 1592

  19. Hot laboratory design on the basis of standardized components

    International Nuclear Information System (INIS)

    Cadrot, J.

    1976-01-01

    The paper describes the principal effects on hot laboratory design brought about over the last 15 years by the use of standardized components developed jointly with the CEA and the industrial associates of AFINE. After a rapid survey of the various advantages of standardization, the author turns to the specific case of a laboratory producing mixed plutonium and uranium oxide fuels, giving a brief description of the glove-boxes and ancillary equipment. He then deals with the design of an isotope production laboratory. The basic component is the DR 200 standard cell, which permits the civil engineering work to be effected on modular principles. Use of a safety-flow pressure regulating valve makes possible pneumatic automation of the production-cell internals. A substantial gain in output is the result. In the next section the paper refers to a pilot facility for irradiated fuel studies, and describes the components used, which require taking into account the high activities and intense radiations encountered in studies of this type. The author then demonstrates the flexibility with which standardized components can be adapted to different uses, thus solving many distinct problems, an example of which is represented by a semi-hot box for handling up to 100g of americium-241. Finally, the paper offers a rapid summary of the effects of standardization at the various stages concerned, from initial design to the commissioning of a hot laboratory. (author)

  20. Exploratory Shaft Seismic Design Basis Working Group report

    International Nuclear Information System (INIS)

    Subramanian, C.V.; King, J.L.; Perkins, D.M.; Mudd, R.W.; Richardson, A.M.; Calovini, J.C.; Van Eeckhout, E.; Emerson, D.O.

    1990-08-01

    This report was prepared for the Yucca Mountain Project (YMP), which is managed by the US Department of Energy. The participants in the YMP are investigating the suitability of a site at Yucca Mountain, Nevada, for construction of a repository for high-level radioactive waste. An exploratory shaft facility (ESF) will be constructed to permit site characterization. The major components of the ESF are two shafts that will be used to provide access to the underground test areas for men, utilities, and ventilation. If a repository is constructed at the site, the exploratory shafts will be converted for use as intake ventilation shafts. In the context of both underground nuclear explosions (conducted at the nearby Nevada Test Site) and earthquakes, the report contains discussions of faulting potential at the site, control motions at depth, material properties of the different rock layers relevant to seismic design, the strain tensor for each of the waveforms along the shaft liners, and the method for combining the different strain components along the shaft liners. The report also describes analytic methods, assumptions used to ensure conservatism, and uncertainties in the data. The analyses show that none of the shafts' structures, systems, or components are important to public radiological safety; therefore, the shafts need only be designed to ensure worker safety, and the report recommends seismic design parameters appropriate for this purpose. 31 refs., 5 figs., 6 tabs

  1. Design basis for the safe disposal of radioactive waste

    International Nuclear Information System (INIS)

    Lewi, J.; Kaluzny, Y.

    1990-01-01

    All radioactive waste disposal sites, regardless of disposal concept, are designed to isolate the radioactive substances contained in such waste for a period at least equal to the time it may remain potentially harmful. Isolation is achieved through the use of containment barriers. This paper summarises the function and limits of different types of barrier used in various disposal systems. For each type of barrier, the paper describes and comments on the site selection criteria and waste packaging requirements applicable in various countries. 13 refs., 1 fig [fr

  2. Improved Design Basis for Laterally Loaded Large Diameter Pile

    DEFF Research Database (Denmark)

    Leth, Caspar Thrane

    of the diameter, depth and soil strength, and increase of each these will give an increase in stiffness. • Cyclic response of a lateral loaded pile is depended on the characteristics of the cyclic load. Behaviour of a monopile is a classic soil-structure interaction problem depending on the pile stiffness....... The target is to improve the use of monopiles as preferred support structure beyond the current limit at a water depth of 30 m. Design of foundations for wind turbines has a large focus on the stiffness of the combined structure, turbine-tower-foundation, which has an influence on the environmental loads...... initial response and a higher ultimate capacity. The initial stiffness of the soil-structure interaction measured in the centrifuge tests, equivalent to initial stiffness of p-y curves, shows a dependency of depth and diameter. Control issues in relation to cyclic tests have resulted in tests...

  3. Targets on the basis of ferrites and high-temperature superconductors for ion-plasma sputtering

    International Nuclear Information System (INIS)

    Lepeshev, A.A.; Saunin, V.N.; Telegin, S.V.; Polyakova, K.P.; Seredkin, V.A.; Pol'skij, A.I.

    2000-01-01

    Paper describes a method to produce targets for ion-plasma sputtering using plasma splaying of the appropriate powders on a cooled metal basis. Application of the plasma process was demonstrated to enable to produce complex shaped targets under the controlled atmosphere on the basis of ceramic materials ensuring their high composition homogeneity, as well as, reliable mechanical and thermal contact of the resultant coating with the base. One carried out experiments in ion-plasma sputtering of targets to prepare ferrite polycrystalline films to be used in magnetooptics and to prepare high-temperature superconductor epitaxial films [ru

  4. 10 CFR 72.94 - Design basis external man-induced events.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Design basis external man-induced events. 72.94 Section 72... WASTE Siting Evaluation Factors § 72.94 Design basis external man-induced events. (a) The region must be examined for both past and present man-made facilities and activities that might endanger the proposed...

  5. 46 CFR 177.310 - Satisfactory service as a design basis.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Satisfactory service as a design basis. 177.310 Section... (UNDER 100 GROSS TONS) CONSTRUCTION AND ARRANGEMENT Hull Structure § 177.310 Satisfactory service as a design basis. When scantlings for the hull, deckhouse, and frames of the vessel differ from those...

  6. Criteria for calculating the efficiency of HEPA filters during and after design basis accidents

    International Nuclear Information System (INIS)

    Bergman, W.; First, M.W.; Anderson, W.L.; Gilbert, H.; Jacox, J.W.

    1994-12-01

    We have reviewed the literature on the performance of high efficiency particulate air (HEPA) filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be structurally damaged and have a residual efficiency of 0%. Despite the many studies on HEPA filter performance under adverse conditions, there are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen when there was insufficient data

  7. Sodium immersible high temperature microphone design description

    International Nuclear Information System (INIS)

    Gavin, A.P.; Anderson, T.T.; Janicek, J.J.

    1975-02-01

    Argonne National Laboratory has developed a rugged high-temperature (HT) microphone for use as a sodium-immersed acoustic monitor in Liquid Metal Fast Breeder Reactors (LMFBRs). Microphones of this design have been extensively tested in room temperature water, in air up to 1200 0 F, and in sodium up to 1200 0 F. They have been successfully installed and employed as acoustic monitors in several operating liquid metal systems. The design, construction sequence, calibration, and testing of these microphones are described. 6 references. (U.S.)

  8. Compensation of temperature frequency pushing in microwave resonator-meters on the basis VCO

    Directory of Open Access Journals (Sweden)

    Drobakhin O. O.

    2008-02-01

    Full Text Available It is shown that the influence of temperature oscillations on the error of measurements of parameters in the case of the application of microwave resonator meters on the basis of a voltage-controlled oscillator (VCO can be minimized by software using a special algorithm of VCO frequency setting correction. An algorithm of VCO frequency setting correction for triangle control voltage is proposed.

  9. Determination of a Basis for Design of a Yam (Dioscorea Spp ...

    African Journals Online (AJOL)

    Manual separation is both tedious and expensive, so the work reported here was done to determine a suitable basis for the design of a mechanical minisett sorter. Results from this study showed that the minisetts cut from the regions of the parent tuber can be separated on the basis of characteristic dimensions of arc length ...

  10. Development of a low-temperature refrigerant on the basis of carbon dioxide; Entwicklung eines Tieftemperaturkaeltemittels auf Basis von Kohlendioxid

    Energy Technology Data Exchange (ETDEWEB)

    Goepfert, Tobias; Hesse, Ullrich [Technische Univ. Dresden (Germany). Bitzer-Stiftungsprofessur fuer Kaelte-, Kryo- und Kompressorentechnik

    2014-07-01

    With the help of known substance properties and derived calculation correlations the mixtures were identified on the basis of R744, which are suitable as a cryogenic refrigerant in the range -50 to -100 degrees Celsius. A prediction and assessment of the suitability of the mixtures is made and the still unknown material properties are shown, which are required for detailed evaluation. [German] Mit Hilfe von bekannten Stoffeigenschaften und abgeleiteten Berechnungskorrelationen wurden Gemische auf der Basis von R744 identifiziert, welche sich als Tieftemperaturkaeltemittel im Bereich von -50 bis -100 Grad Celsius eigenen. Es wird eine Prognose und Abschaetzung der Eignung der Stoffgemische vorgenommen und es werden die noch unbekannten Stoffeigenschaften dargestellt, die zur genauen Bewertung benoetigt werden.

  11. An Improved Setpoint Determination Methodology for the Plant Protection System Considering Beyond Design Basis Events

    International Nuclear Information System (INIS)

    Lee, C.J.; Baik, K.I.; Baek, S.M.; Park, K.-M.; Lee, S.J.

    2013-06-01

    According to the nuclear regulations and industry standards, the trip setpoint and allowable value for the plant protection system have been determined by considering design basis events. In order to improve the safety of a nuclear power plant, an attempt has been made to develop an improved setpoint determination methodology for the plant protection system trip parameter considering not only a design basis event but also a beyond design basis event. The results of a quantitative evaluation performed for the Advanced Power Reactor 1400 nuclear power plant in Korea are presented herein. The results confirmed that the proposed methodology is able to improve the nuclear power plant's safety by determining more reasonable setpoints that can cover beyond design basis events. (authors)

  12. Site selection and design basis of the National Disposal Facility for LILW. Geological and engineering barriers

    International Nuclear Information System (INIS)

    Boyanov, S.

    2010-01-01

    Content of the presentation: Site selection; Characteristics of the “Radiana” site (location, geological structure, physical and mechanical properties, hydro-geological conditions); Design basis of the Disposal Facility; Migration analysis; Safety assessment approach

  13. The Mixed Waste Management Facility. Design basis integrated operations plan (Title I design)

    International Nuclear Information System (INIS)

    1994-12-01

    The Mixed Waste Management Facility (MWMF) will be a fully integrated, pilotscale facility for the demonstration of low-level, organic-matrix mixed waste treatment technologies. It will provide the bridge from bench-scale demonstrated technologies to the deployment and operation of full-scale treatment facilities. The MWMF is a key element in reducing the risk in deployment of effective and environmentally acceptable treatment processes for organic mixed-waste streams. The MWMF will provide the engineering test data, formal evaluation, and operating experience that will be required for these demonstration systems to become accepted by EPA and deployable in waste treatment facilities. The deployment will also demonstrate how to approach the permitting process with the regulatory agencies and how to operate and maintain the processes in a safe manner. This document describes, at a high level, how the facility will be designed and operated to achieve this mission. It frequently refers the reader to additional documentation that provides more detail in specific areas. Effective evaluation of a technology consists of a variety of informal and formal demonstrations involving individual technology systems or subsystems, integrated technology system combinations, or complete integrated treatment trains. Informal demonstrations will typically be used to gather general operating information and to establish a basis for development of formal demonstration plans. Formal demonstrations consist of a specific series of tests that are used to rigorously demonstrate the operation or performance of a specific system configuration

  14. Guidance on the implementation of modifications to mitigate beyond design basis accidents

    International Nuclear Information System (INIS)

    Harris, S.; Marczak, J.; O'Neill, M.

    2014-01-01

    Following the events at Fukushima, Canadian Nuclear Power Plants (NPP) procured equipment and initiated modifications to improve response capability for Beyond Design Basis Accidents (BDBA). These changes were not typical of other design modifications to the nuclear power plants and reinforced the need for additional guidance for modifications to address BDBA. This paper describes the guidance that was developed to guide the design, procurement, installation, operation, and maintenance of equipment and modifications to mitigate BDBAs. The guidance developed prescribes a graded approach based on a categorization of the nature of the modification. Four categories of modifications are introduced, with the distinction being the degree of interface with existing design basis systems, structures and components (SSCs). This has resulted in a cost-effective means of implementing additional capability to mitigate BDBA conditions, and yet ensure the design basis capability of SSCs is maintained. Operating experience with use of the guidance is also discussed. (author)

  15. Guidance on the implementation of modifications to mitigate beyond design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    Harris, S.; Marczak, J.; O' Neill, M. [Ontario Power Generation, Pickering, ON (Canada)

    2014-07-01

    Following the events at Fukushima, Canadian Nuclear Power Plants (NPP) procured equipment and initiated modifications to improve response capability for Beyond Design Basis Accidents (BDBA). These changes were not typical of other design modifications to the nuclear power plants and reinforced the need for additional guidance for modifications to address BDBA. This paper describes the guidance that was developed to guide the design, procurement, installation, operation, and maintenance of equipment and modifications to mitigate BDBAs. The guidance developed prescribes a graded approach based on a categorization of the nature of the modification. Four categories of modifications are introduced, with the distinction being the degree of interface with existing design basis systems, structures and components (SSCs). This has resulted in a cost-effective means of implementing additional capability to mitigate BDBA conditions, and yet ensure the design basis capability of SSCs is maintained. Operating experience with use of the guidance is also discussed. (author)

  16. Guidance on the Implementation of Modifications to Mitigate Beyond Design Basis Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Dermarkar, F.; Marczak, J.; O’Neill, M., E-mail: fred.dermarkar@opg.com [Ontario Power Generation, Pickering, Ontario (Canada)

    2014-10-15

    Following the events at Fukushima, Canadian Nuclear Power Plants (NPP) procured equipment and initiated modifications to improve response capability for Beyond Design Basis Accidents (BDBA). These changes were not typical of other design modifications to the nuclear power plants and reinforced the need for additional guidance for modifications to address BDBA. This paper describes the guidance that was developed to guide the design, procurement, installation, operation, and maintenance of equipment and modifications to mitigate BDBAs. The guidance developed prescribes a graded approach based on a categorization of the nature of the modification. Four categories of modifications are introduced, with the distinction being the degree of interface with existing design basis systems, structures and components (SSCs). This has resulted in a cost-effective means of implementing additional capability to mitigate BDBA conditions, and yet ensure the design basis capability of SSCs is maintained. (author)

  17. A Designed Room Temperature Multilayered Magnetic Semiconductor

    Science.gov (United States)

    Bouma, Dinah Simone; Charilaou, Michalis; Bordel, Catherine; Duchin, Ryan; Barriga, Alexander; Farmer, Adam; Hellman, Frances; Materials Science Division, Lawrence Berkeley National Lab Team

    2015-03-01

    A room temperature magnetic semiconductor has been designed and fabricated by using an epitaxial antiferromagnet (NiO) grown in the (111) orientation, which gives surface uncompensated magnetism for an odd number of planes, layered with the lightly doped semiconductor Al-doped ZnO (AZO). Magnetization and Hall effect measurements of multilayers of NiO and AZO are presented for varying thickness of each. The magnetic properties vary as a function of the number of Ni planes in each NiO layer; an odd number of Ni planes yields on each NiO layer an uncompensated moment which is RKKY-coupled to the moments on adjacent NiO layers via the carriers in the AZO. This RKKY coupling oscillates with the AZO layer thickness, and it disappears entirely in samples where the AZO is replaced with undoped ZnO. The anomalous Hall effect data indicate that the carriers in the AZO are spin-polarized according to the direction of the applied field at both low temperature and room temperature. NiO/AZO multilayers are therefore a promising candidate for spintronic applications demanding a room-temperature semiconductor.

  18. Establishing design basis threats for the physical protection of nuclear materials and facilities

    International Nuclear Information System (INIS)

    Chetvergov, S.

    2001-01-01

    In the area of nuclear energy utilization, the Republic of Kazakhstan follows the standards of international legislation and is a participant of the Nuclear Weapons Non-proliferation Treaty as a country that does not have nuclear weapons. In the framework of this treaty, Kazakhstan provides for the measures to ensure the regime of nonproliferation. The Republic signed the Agreement with the IAEA on the guarantee that was ratified by the Presidential Decree in 1995. Now the Government of the RK is considering the Convention on Physical Protection of Nuclear Materials. Kazakhstan legislation in the area of nuclear energy utilization is represented by a set of laws: the main of them is the Law of the Republic of Kazakhstan 'On the utilization of atomic energy', dated April 14, 1997. According to the Law, the issues of physical protection are regulated by interdepartmental guideline documents. Nuclear science and industry of RK include: Enterprises on uranium mining and processing; Ulba metallurgical plant, manufacturing fuel pellets of uranium dioxide for heat release assemblies of RBMK and WWR reactor types, with the enrichment on U235 1.6-4.4%; Power plant in Aktau for heat and power supply and water desalination, based on fast breeder reactor BN-350; Research reactors of National Nuclear Center: WWR-K - water-water reactor, with 10 MW power, uses highly enriched uranium (up to 36% of U-235); IVG.1M - water-water heterogeneous reactor of vessel type on thermal neutrons, maximum power is 35 MW; IGR - impulse homogeneous graphite reactor on thermal neutrons, with graphite reflector; RA - high temperature gas cooled reactor on thermal neutrons, 0.5 MW power. The establishment of design basis threats for nuclear objects in the Republic of Kazakhstan is an urgent problem because of the developing military-political situation in the region. It is necessary to specify important elements affecting the specific features of the design basis threat: military operations of

  19. Some conditions affecting the definition of design basis accidents relating to sodium/water reactions

    International Nuclear Information System (INIS)

    Bolt, P.R.

    1984-01-01

    The possible damaging effects of large sodium/water reactions on the steam generator, IHX and secondary circuit are considered. The conditions to be considered in defining the design basis accidents for these components are discussed, together with some of the assumptions that may be associated with design assessments of the scale of the accidents. (author)

  20. Multi dimensional analysis of Design Basis Events using MARS-LMR

    International Nuclear Information System (INIS)

    Woo, Seung Min; Chang, Soon Heung

    2012-01-01

    Highlights: ► The one dimensional analyzed sodium hot pool is modified to a three dimensional node system, because the one dimensional analysis cannot represent the phenomena of the inside pool of a big size pool with many compositions. ► The results of the multi-dimensional analysis compared with the one dimensional analysis results in normal operation, TOP (Transient of Over Power), LOF (Loss of Flow), and LOHS (Loss of Heat Sink) conditions. ► The difference of the sodium flow pattern due to structure effect in the hot pool and mass flow rates in the core lead the different sodium temperature and temperature history under transient condition. - Abstract: KALIMER-600 (Korea Advanced Liquid Metal Reactor), which is a pool type SFR (Sodium-cooled Fast Reactor), was developed by KAERI (Korea Atomic Energy Research Institute). DBE (Design Basis Events) for KALIMER-600 has been analyzed in the one dimension. In this study, the one dimensional analyzed sodium hot pool is modified to a three dimensional node system, because the one dimensional analysis cannot represent the phenomena of the inside pool of a big size pool with many compositions, such as UIS (Upper Internal Structure), IHX (Intermediate Heat eXchanger), DHX (Decay Heat eXchanger), and pump. The results of the multi-dimensional analysis compared with the one dimensional analysis results in normal operation, TOP (Transient of Over Power), LOF (Loss of Flow), and LOHS (Loss of Heat Sink) conditions. First, the results in normal operation condition show the good agreement between the one and multi-dimensional analysis. However, according to the sodium temperatures of the core inlet, outlet, the fuel central line, cladding and PDRC (Passive Decay heat Removal Circuit), the temperatures of the one dimensional analysis are generally higher than the multi-dimensional analysis in conditions except the normal operation state, and the PDRC operation time in the one dimensional analysis is generally longer than

  1. High temperature creep-fatigue design

    International Nuclear Information System (INIS)

    Tavassoli, A. A. F.; Fournier, B.; Sauzay, M.

    2010-01-01

    Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowable and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed. (authors)

  2. High temperature creep-fatigue design

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A. A. F.; Fournier, B.; Sauzay, M. [CEA Saclay, DEN DMN, F-91191 Gif Sur Yvette (France)

    2010-07-01

    Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowable and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed. (authors)

  3. Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

    International Nuclear Information System (INIS)

    Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

    2010-01-01

    To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

  4. Sea Surface Temperature Modeling using Radial Basis Function Networks With a Dynamically Weighted Particle Filter

    KAUST Repository

    Ryu, Duchwan

    2013-03-01

    The sea surface temperature (SST) is an important factor of the earth climate system. A deep understanding of SST is essential for climate monitoring and prediction. In general, SST follows a nonlinear pattern in both time and location and can be modeled by a dynamic system which changes with time and location. In this article, we propose a radial basis function network-based dynamic model which is able to catch the nonlinearity of the data and propose to use the dynamically weighted particle filter to estimate the parameters of the dynamic model. We analyze the SST observed in the Caribbean Islands area after a hurricane using the proposed dynamic model. Comparing to the traditional grid-based approach that requires a supercomputer due to its high computational demand, our approach requires much less CPU time and makes real-time forecasting of SST doable on a personal computer. Supplementary materials for this article are available online. © 2013 American Statistical Association.

  5. Defense-in-depth approach against a beyond design basis event

    Energy Technology Data Exchange (ETDEWEB)

    Hoang, H., E-mail: Hoa.hoang@ge.com [GE Hitachi Nuclear Energy, 1989 Little Orchard St., 95125 San Jose, California (United States)

    2013-10-15

    The US industry, with the approval of the Nuclear Regulatory Commission, is promoting an approach to add diverse and flexible mitigation strategies, or Flex, that will increase the defense-in-depth capability for the nuclear power plants in the event of beyond design basis event, such as at the Fukushima Dai-ichi station. The objective of Flex is to establish and indefinite coping capability to prevent damage to the fuel in the core and spent fuel pool, and to maintain the containment function by utilizing installed equipment, on-site portable equipment and pre-staged off-site resources. This capability will address both an extended loss of all Ac power and a loss of ultimate heat sink which could arise following a design basis event with additional failures, and conditions from a beyond design basis event. (author)

  6. Defense-in-depth approach against a beyond design basis event

    International Nuclear Information System (INIS)

    Hoang, H.

    2013-10-01

    The US industry, with the approval of the Nuclear Regulatory Commission, is promoting an approach to add diverse and flexible mitigation strategies, or Flex, that will increase the defense-in-depth capability for the nuclear power plants in the event of beyond design basis event, such as at the Fukushima Dai-ichi station. The objective of Flex is to establish and indefinite coping capability to prevent damage to the fuel in the core and spent fuel pool, and to maintain the containment function by utilizing installed equipment, on-site portable equipment and pre-staged off-site resources. This capability will address both an extended loss of all Ac power and a loss of ultimate heat sink which could arise following a design basis event with additional failures, and conditions from a beyond design basis event. (author)

  7. Design Basis Threat (DBT) Approach for the First NPP Security System in Indonesia

    International Nuclear Information System (INIS)

    Ign Djoko Irianto

    2004-01-01

    Design Basis Threat (DBT) is one of the main factors to be taken into account in the design of physical protection system of nuclear facility. In accordance with IAEA's recommendations outlined in INFCIRC/225/Rev.4 (Corrected), DBT is defined as: attributes and characteristics of potential insider and/or external adversaries, who might attempt unauthorized removal of nuclear material or sabotage against the nuclear facilities. There are three types of adversary that must be considered in DBT, such as adversary who comes from the outside (external adversary), adversary who comes from the inside (internal adversary), and adversary who comes from outside and colludes with insiders. Current situation in Indonesia, where many bomb attacks occurred, requires serious attention on DBT in the physical protection design of NPP which is to be built in Indonesia. This paper is intended to describe the methodology on how to create and implement a Design Basis Threat in the design process of NPP physical protection in Indonesia. (author)

  8. Prediction of water formation temperature in natural gas dehydrators using radial basis function (RBF neural networks

    Directory of Open Access Journals (Sweden)

    Tatar Afshin

    2016-03-01

    Full Text Available Raw natural gases usually contain water. It is very important to remove the water from these gases through dehydration processes due to economic reasons and safety considerations. One of the most important methods for water removal from these gases is using dehydration units which use Triethylene glycol (TEG. The TEG concentration at which all water is removed and dew point characteristics of mixture are two important parameters, which should be taken into account in TEG dehydration system. Hence, developing a reliable and accurate model to predict the performance of such a system seems to be very important in gas engineering operations. This study highlights the use of intelligent modeling techniques such as Multilayer perceptron (MLP and Radial Basis Function Neural Network (RBF-ANN to predict the equilibrium water dew point in a stream of natural gas based on the TEG concentration of stream and contractor temperature. Literature data set used in this study covers temperatures from 10 °C to 80 °C and TEG concentrations from 90.000% to 99.999%. Results showed that both models are accurate in prediction of experimental data and the MLP model gives more accurate predictions compared to RBF model.

  9. Assessment Of Source Term And Radiological Consequences For Design Basis Accident And Beyond Design Basis Accident Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong; Tran Tri Vien

    2011-01-01

    The paper presents results of the assessment of source terms and radiological consequences for the Design Basis Accident (DBA) and Beyond Design Basis Accident (BDBA) of the Dalat Nuclear Research Reactor. The dropping of one fuel assembly during fuel handling operation leading to the failure of fuel cladding and the release of fission products into the environment was selected as a DBA for the analysis. For the BDBA, the introduction of a step positive reactivity due to the falling of a heavy block from the rotating bridge crane in the reactor hall onto a part of the platform where are disposed the control rod drives is postulated. The result of the radiological consequence analyses shows that doses to members of the public are below annual dose limit for both DBA and BDBA events. However, doses from exposure to operating staff and experimenters working inside the reactor hall are predicted to be very high in case of BDBA and therefore the protective actions should be taken when the accident occurs. (author)

  10. Analysis of regulatory requirement for beyond design basis events of SMART

    International Nuclear Information System (INIS)

    Kim, W. S.; Seol, K. W.

    2000-01-01

    To enhance the safety of SMART reactor, safety and regulatory requirements associated with beyond design basis events (beyond BDE), which were developed and applied to advanced light water reactor designs, were analyzed along with a design status of passive reactor. And, based on these requirements, their applicability on the SMART design was evaluated. In the design aspect, severe accident prevention and mitigation features, containment performance, and accident management were analyzed. The evaluation results show that the requirement related to beyond DBE such as ATWS, loss of residual heat removal during shutdown operation, station blackout, fire, inter-system LOCA, and well-known events from severe accident phenomena is applicable to the SMART design. However, comprehensive approach against beyond DBE is not yet provided in the SMART design, and then it is required to designate and analyze the beyond DBE-related features. This study is expected to contribute to efforts to improve plant safety and to establish regulatory requirements for safety review

  11. Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, Rajiv; Hibler, Lyle F.; Coleman, Andre M.; Ward, Duane L.

    2011-11-01

    The purpose of this document is to describe approaches and methods for estimation of the design-basis flood at nuclear power plant sites. Chapter 1 defines the design-basis flood and lists the U.S. Nuclear Regulatory Commission's (NRC) regulations that require estimation of the design-basis flood. For comparison, the design-basis flood estimation methods used by other Federal agencies are also described. A brief discussion of the recommendations of the International Atomic Energy Agency for estimation of the design-basis floods in its member States is also included.

  12. Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America

    International Nuclear Information System (INIS)

    Prasad, Rajiv; Hibler, Lyle F.; Coleman, Andre M.; Ward, Duane L.

    2011-01-01

    The purpose of this document is to describe approaches and methods for estimation of the design-basis flood at nuclear power plant sites. Chapter 1 defines the design-basis flood and lists the U.S. Nuclear Regulatory Commission's (NRC) regulations that require estimation of the design-basis flood. For comparison, the design-basis flood estimation methods used by other Federal agencies are also described. A brief discussion of the recommendations of the International Atomic Energy Agency for estimation of the design-basis floods in its member States is also included.

  13. Study on effective prestressing effects on concrete containment under the design-basis pressure condition

    International Nuclear Information System (INIS)

    Sun Feng; Pan Rong; Wang Lu; Mao Huan; Yang Yu

    2013-01-01

    Prestressing technology is widely used in nuclear power plant containment building, and the durability of containment structure is affected directly by the distribution and loss of prestressing value under design-basis pressure. Containment structure and the distribution of prestressing system are introduced briefly. Furthermore, the calculating process of horizontal prestressing bunch loss near the equipment hatch hole is put forward in details, and the containment structure prestressing loss when 5-year pressure test is obtained. Based above analysis, the finite element model of the prestressed concrete containment structure is built by using ANSYS code, the prestressing effect on concrete containment is analysed. The results show that most of the design pressure is bore by the prestressing system under the design-basis pressure, so the containment structure is safe. These conclusions are consistent with prestressing containment system design concepts, which can provide reference to the engineering staff. (authors)

  14. The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

    Energy Technology Data Exchange (ETDEWEB)

    Williams, M. G.; Mouser, M. R.; Simon, J. B. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been

  15. The power of simplification: Operator interface with the AP1000R during design-basis and beyond design-basis events

    International Nuclear Information System (INIS)

    Williams, M. G.; Mouser, M. R.; Simon, J. B.

    2012-01-01

    The AP1000 R plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been designed

  16. Methods for very high temperature design

    International Nuclear Information System (INIS)

    Blass, J.J.; Corum, J.M.; Chang, S.J.

    1989-01-01

    Design rules and procedures for high-temperature, gas-cooled reactor components are being formulated as an ASME Boiler and Pressure Vessel Code Case. A draft of the Case, patterned after Code Case N-47, and limited to Inconel 617 and temperatures of 982/degree/C (1800/degree/F) or less, will be completed in 1989 for consideration by relevant Code committees. The purpose of this paper is to provide a synopsis of the significant differences between the draft Case and N-47, and to provide more complete accounts of the development of allowable stress and stress rupture values and the development of isochronous stress vs strain curves, in both of which Oak Ridge National Laboratory (ORNL) played a principal role. The isochronous curves, which represent average behavior for many heats of Inconel 617, were based in part on a unified constitutive model developed at ORNL. Details are also provided of this model of inelastic deformation behavior, which does not distinguish between rate-dependent plasticity and time-dependent creep, along with comparisons between calculated and observed results of tests conducted on a typical heat of Inconel 617 by the General Electric Company for the Department of Energy. 4 refs., 15 figs., 1 tab

  17. Reconfigurable Flight Control Design using a Robust Servo LQR and Radial Basis Function Neural Networks

    Science.gov (United States)

    Burken, John J.

    2005-01-01

    This viewgraph presentation reviews the use of a Robust Servo Linear Quadratic Regulator (LQR) and a Radial Basis Function (RBF) Neural Network in reconfigurable flight control designs in adaptation to a aircraft part failure. The method uses a robust LQR servomechanism design with model Reference adaptive control, and RBF neural networks. During the failure the LQR servomechanism behaved well, and using the neural networks improved the tracking.

  18. Chemical data for the calculation of fission product releases in design basis faults in PWRs

    International Nuclear Information System (INIS)

    Ali, S.M.; Bawden, R.J.; Garbett, K.; Deane, A.M.; Large, N.R.

    1982-04-01

    This review considers the chemistry of caesium and iodine and their volatility under the conditions which would exist during a number of design-basis faults. It recommends values which should be used for the distribution of these elements between liquid and gas phases. (author)

  19. Beyond-design-basis accident management in the RF regulation documents

    International Nuclear Information System (INIS)

    Bukrinskij, A.M.

    2010-01-01

    The article observes the issues of the management of beyond-design-basis accidents (BDBA) in the existing regulations in Russia. The ideology of the approach to the definition of the BDBA list to formulate the management guidelines has been proposed [ru

  20. NDARC-NASA Design and Analysis of Rotorcraft Theoretical Basis and Architecture

    Science.gov (United States)

    Johnson, Wayne

    2010-01-01

    The theoretical basis and architecture of the conceptual design tool NDARC (NASA Design and Analysis of Rotorcraft) are described. The principal tasks of NDARC are to design (or size) a rotorcraft to satisfy specified design conditions and missions, and then analyze the performance of the aircraft for a set of off-design missions and point operating conditions. The aircraft consists of a set of components, including fuselage, rotors, wings, tails, and propulsion. For each component, attributes such as performance, drag, and weight can be calculated. The aircraft attributes are obtained from the sum of the component attributes. NDARC provides a capability to model general rotorcraft configurations, and estimate the performance and attributes of advanced rotor concepts. The software has been implemented with low-fidelity models, typical of the conceptual design environment. Incorporation of higher-fidelity models will be possible, as the architecture of the code accommodates configuration flexibility, a hierarchy of models, and ultimately multidisciplinary design, analysis and optimization.

  1. Radial basis function (RBF) neural network control for mechanical systems design, analysis and Matlab simulation

    CERN Document Server

    Liu, Jinkun

    2013-01-01

    Radial Basis Function (RBF) Neural Network Control for Mechanical Systems is motivated by the need for systematic design approaches to stable adaptive control system design using neural network approximation-based techniques. The main objectives of the book are to introduce the concrete design methods and MATLAB simulation of stable adaptive RBF neural control strategies. In this book, a broad range of implementable neural network control design methods for mechanical systems are presented, such as robot manipulators, inverted pendulums, single link flexible joint robots, motors, etc. Advanced neural network controller design methods and their stability analysis are explored. The book provides readers with the fundamentals of neural network control system design.   This book is intended for the researchers in the fields of neural adaptive control, mechanical systems, Matlab simulation, engineering design, robotics and automation. Jinkun Liu is a professor at Beijing University of Aeronautics and Astronauti...

  2. 77 FR 64564 - Implementation of Regulatory Guide 1.221 on Design-Basis Hurricane and Hurricane Missiles

    Science.gov (United States)

    2012-10-22

    ...-Basis Hurricane and Hurricane Missiles AGENCY: Nuclear Regulatory Commission. ACTION: Proposed interim...-ISG-024, ``Implementation of Regulatory Guide 1.221 on Design-Basis Hurricane and Hurricane Missiles....221, ``Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants.'' DATES: Submit...

  3. Design Basis Provisions for New and Existing Nuclear Power Plants and Nuclear Fuel Cycle Facilities in India

    International Nuclear Information System (INIS)

    Soni, R.S.

    2013-01-01

    India has 3-Stage Nuclear Power Program. • Various facilities under design, construction or operation. • Design Basis Knowledge Management (DBKM) is an important and challenging task. • Design Basis Knowledge contributes towards: - Safe operation of running plants; - Design and construction of new facilities; - Addresses issues related to future decommissioning activities

  4. Reducing Production Basis Risk through Rainfall Intensity Frequency (RIF) Indexes: Global Sensitivity Analysis' Implication on Policy Design

    Science.gov (United States)

    Muneepeerakul, Chitsomanus; Huffaker, Ray; Munoz-Carpena, Rafael

    2016-04-01

    The weather index insurance promises financial resilience to farmers struck by harsh weather conditions with swift compensation at affordable premium thanks to its minimal adverse selection and moral hazard. Despite these advantages, the very nature of indexing causes the presence of "production basis risk" that the selected weather indexes and their thresholds do not correspond to actual damages. To reduce basis risk without additional data collection cost, we propose the use of rain intensity and frequency as indexes as it could offer better protection at the lower premium by avoiding basis risk-strike trade-off inherent in the total rainfall index. We present empirical evidences and modeling results that even under the similar cumulative rainfall and temperature environment, yield can significantly differ especially for drought sensitive crops. We further show that deriving the trigger level and payoff function from regression between historical yield and total rainfall data may pose significant basis risk owing to their non-unique relationship in the insured range of rainfall. Lastly, we discuss the design of index insurance in terms of contract specifications based on the results from global sensitivity analysis.

  5. NPP Design Basis Handover and Knowledge Preservation from Subcontractors, Vendors and EPC

    International Nuclear Information System (INIS)

    Freeland, Kent

    2013-01-01

    Using PLM-based Workflow for Configuration Management (CM) in the Nuclear Power Industry Advantages – some work to do! • NPP’s must adapt to using PLM-based solutions to support CM and to synchronize design changes to asset or product changes, and reduce “slipstreaming”. In the NPP world, this often appears as events that circumvent CM – for example, non-approved parts substitutions and “temporary” plant modifications that are never removed. • PLM serves as the method for unifying the application of requirements to design changes, processes and workflow. In NPP’s, requirements are generally considered only relevant to designs – not process and workflow. • PLM supports Configuration Management and Design Basis in Regulator Action Tracking for NPP’s, and application of PLM-based CM to regulator action and compliance systems. This is a poorly-understood application of CM in NPP’s, yet these elements control large parts of the NPP design basis. • Suppliers, EPC’s and Technology Vendors must also understand the role of CM, SE and PLM in construction of new standards-driven NPP designs (like EPR and Westinghouse AP-1000 NPP designs), as well as understanding the role and handling of Knowledge Systems

  6. Development of a design basis tornado and structural design criteria for Lawrence Livermore Laboratory's Site 300

    International Nuclear Information System (INIS)

    McDonald, J.R.; Minor, J.E.; Mehta, K.C.

    1975-11-01

    Criteria are prescribed and guidance is provided for professional personnel who are involved with the evaluation of existing buildings and facilities at Site 300 near Livermore, California to resist the possible effects of extreme winds and tornadoes. The development of parameters for the effects of tornadoes and extreme winds and guidelines for evaluation and design of structures are presented. The investigations conducted are summarized and the techniques used for arriving at the combined tornado and extreme wind risk model are discussed. The guidelines for structural design methods for calculating pressure distributions on walls and roofs of structures and methods for accommodating impact loads from missiles are also presented

  7. Development and comparision of techniques for estimating design basis flood flows for nuclear power plants

    International Nuclear Information System (INIS)

    1980-05-01

    Estimation of the design basis flood for Nuclear Power Plants can be carried out using either deterministic or stochastic techniques. Stochastic techniques, while widely used for the solution of a variety of hydrological and other problems, have not been used to date (1980) in connection with the estimation of design basis flood for NPP siting. This study compares the two techniques against one specific river site (Galt on the Grand River, Ontario). The study concludes that both techniques lead to comparable results , but that stochastic techniques have the advantage of extracting maximum information from available data and presenting the results (flood flow) as a continuous function of probability together with estimation of confidence limits. (author)

  8. Ground motion following selection of SRS design basis earthquake and associated deterministic approach

    International Nuclear Information System (INIS)

    1991-03-01

    This report summarizes the results of a deterministic assessment of earthquake ground motions at the Savannah River Site (SRS). The purpose of this study is to assist the Environmental Sciences Section of the Savannah River Laboratory in reevaluating the design basis earthquake (DBE) ground motion at SRS during approaches defined in Appendix A to 10 CFR Part 100. This work is in support of the Seismic Engineering Section's Seismic Qualification Program for reactor restart

  9. Interregional Knowledge Management Workshop on Life Cycle Management of Design Basis Information. Issues, Challenges, Approaches

    International Nuclear Information System (INIS)

    Šula, Radek

    2013-01-01

    Introduction and objectives: • It is evident that the design basis area is from the point of view of knowledge sharing extremely complicated. • Time is changing and puts on us ever greater demands. • We have to analyze the near and remote surroundings and have to simplified the problem of knowledge sharing in that area. • I believe that it is graspable task for knowledge management and I will try to outline some possible context and approaches

  10. Discussion about design basis flood of site of research reactors by river

    International Nuclear Information System (INIS)

    Rong Feng; Zhao Jianjun; Du Qiaomin; Zhang Lingyan

    2006-01-01

    This paper presents the well-defined standard in relation to design the basis flood of the sites of research reactors by river. It is based on the concept of some relational standards, analysis of hydrological calculation technology and methods, and analysis of accident dangerous degrees of research reactor, as well as in combination with the engineering practices. The flood preventing standard for research reactors with higher power should be the same with that of the nuclear power plants. (authors)

  11. The Swedish Utilities joint approach to form common basis for design requirements for the future

    International Nuclear Information System (INIS)

    Hansson, B.

    1998-01-01

    The Owners of the Swedish Nuclear Power Plants have decided to form a document that should state the design principals and requirement for cost-effective and continuous development of the reactor safety in the future. The development of this document will be a part of the modernization and development of the Swedish Nuclear Power Plants. The basis for this document is an evaluation of Swedish and International standards and regulations as IAEA/INSAG, US-regulations, EUR etc. (author)

  12. Design and Modeling of RF Power Amplifiers with Radial Basis Function Artificial Neural Networks

    OpenAIRE

    Ali Reza Zirak; Sobhan Roshani

    2016-01-01

    A radial basis function (RBF) artificial neural network model for a designed high efficiency radio frequency class-F power amplifier (PA) is presented in this paper. The presented amplifier is designed at 1.8 GHz operating frequency with 12 dB of gain and 36 dBm of 1dB output compression point. The obtained power added efficiency (PAE) for the presented PA is 76% under 26 dBm input power. The proposed RBF model uses input and DC power of the PA as inputs variables and considers output power a...

  13. NSSEFF Designing New Higher Temperature Superconductors

    Science.gov (United States)

    2017-04-13

    AFRL-AFOSR-VA-TR-2017-0083 NSSEFF - DESIGINING NEW HIGHER TEMPERATURE SUPERCONDUCTORS Meigan Aronson THE RESEARCH FOUNDATION OF STATE UNIVERSITY OF...2015 4. TITLE AND SUBTITLE NSSEFF - DESIGINING NEW HIGHER TEMPERATURE SUPERCONDUCTORS 5a.  CONTRACT NUMBER 5b.  GRANT NUMBER FA9550-10-1-0191 5c...materials, identifying the most promising candidates. 15. SUBJECT TERMS TEMPERATURE, SUPERCONDUCTOR 16. SECURITY CLASSIFICATION OF: 17. LIMITATION OF

  14. Design basis for creep of zirconium alloy components in a fast neutron flux

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Fidleris, V.

    1975-01-01

    The chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behaviour of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

  15. Design basis for creep of zirconium alloy components in a fast neutron flux

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Fidleris, V.

    1974-01-01

    The Chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behavior of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time, and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

  16. Characterisation of Liquefaction Effects for Beyond-Design Basis Safety Assessment of Nuclear Power Plants

    Science.gov (United States)

    Bán, Zoltán; Győri, Erzsébet; János Katona, Tamás; Tóth, László

    2015-04-01

    Preparedness of nuclear power plants to beyond design base external effects became high importance after 11th of March 2011 Great Tohoku Earthquakes. In case of some nuclear power plants constructed at the soft soil sites, liquefaction should be considered as a beyond design basis hazard. The consequences of liquefaction have to be analysed with the aim of definition of post-event plant condition, identification of plant vulnerabilities and planning the necessary measures for accident management. In the paper, the methodology of the analysis of liquefaction effects for nuclear power plants is outlined. The case of Nuclear Power Plant at Paks, Hungary is used as an example for demonstration of practical importance of the presented results and considerations. Contrary to the design, conservatism of the methodology for the evaluation of beyond design basis liquefaction effects for an operating plant has to be limited to a reasonable level. Consequently, applicability of all existing methods has to be considered for the best estimation. The adequacy and conclusiveness of the results is mainly limited by the epistemic uncertainty of the methods used for liquefaction hazard definition and definition of engineering parameters characterizing the consequences of liquefaction. The methods have to comply with controversial requirements. They have to be consistent and widely accepted and used in the practice. They have to be based on the comprehensive database. They have to provide basis for the evaluation of dominating engineering parameters that control the post-liquefaction response of the plant structures. Experience of Kashiwazaki-Kariwa plant hit by Niigata-ken Chuetsu-oki earthquake of 16 July 2007 and analysis of site conditions and plant layout at Paks plant have shown that the differential settlement is found to be the dominating effect in case considered. They have to be based on the probabilistic seismic hazard assessment and allow the integration into logic

  17. Design methods for high temperature power plant structures

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1984-01-01

    The subject is discussed under the headings: introduction (scope of paper - reviews of design methods and design criteria currently in use for both nuclear and fossil fuelled power plant; examples chosen are (a) BS 1113, representative of design codes employed for power station boiler plant; (b) ASME Code Case N47, which is being developed for high temperature nuclear reactors, especially the liquid metal fast breeder reactor); design codes for power station boilers; Code Case N47 (design in the absence of thermal shock and thermal fatigue; design against cyclic loading at high temperature; further research in support of high temperature design methods and criteria for LMFBRs); concluding remarks. (U.K.)

  18. Structural analysis of the CAREM-25 nuclear power plant subjected to the design basis accident and seismic loads

    International Nuclear Information System (INIS)

    Ambrosini, Daniel; Codina, Ramón H.; Curadelli, Oscar; Martínez, Carlos A.

    2017-01-01

    Highlights: • Structural analysis of CAREM-25 NPP is presented. • Full 3D numerical model was developed. • Transient thermal and static structural analyses were performed. • Modeling guidelines for numerical structural analysis of NPP are recommended. • Envelope condition of DBA dominates the structural behavior. - Abstract: In this paper, a numerical study about the structural response of the Argentine nuclear power plant CAREM-25 subjected to the design basis accident (DBA) and seismic loads is presented. Taking into account the hardware capabilities available, a full 3D finite element model was adopted. A significant part of the building was modeled using more than 2 M solid elements. In order to take into account the foundation flexibility, linear springs were used. The springs and the model were calibrated against a greater model used to study the soil-structure interaction. The structure was subjected to the DBA and seismic loads as combinations defined by ASME international code. First, a transient thermal analysis was performed with the conditions defined by DBA and evaluating the time history of the temperature of the model, each 1 h until 36 h. The final results of this stage were considered as initial conditions of a static structural analysis including the pressure defined by DBA. Finally, an equivalent static analysis was performed to analyze the seismic response considering the design basis spectra for the site. The different loads were combined and the abnormal/extreme environmental combination was the most unfavorable for the structure, defining the design.

  19. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    International Nuclear Information System (INIS)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W.

    2016-01-01

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  20. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2016-05-15

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  1. E-learning course: Basis of Harvest and Preservation of Tissues – design and initial experience

    Directory of Open Access Journals (Sweden)

    Pavel Měřička

    2014-05-01

    Full Text Available Background: The design and initial experience with the e-learning course “Basis of Harvest and Preservation of Tissues” used as a support of an elective subject is presented. The aim of the e-learning course was to enable the students to learn the theoretical principles of the subject individually and to present the gained knowledge at the final seminar. Methods: All functions of the course were operated in Moodle, local application of the Charles University in Prague, Faculty of Medicine in Hradec Králové. The course was divided into 3 main topics corresponding with topics of lectures: 1. Principles of tissue and organ donation, 2. Low temperature preservation of cells, tissues and organs, 3. Quality and safety assurance in practice of tissue and procurement establishments. A test consisting of 5 questions selected randomly from the bank of questions followed each topic. If the student answers correctly at least 3 questions he is allowed to pass to the next topic. The fourth topic “Basic processes in the tissue establishment and principles of their validation” was added into the electronic version as a tool for repeating and improving of knowledge. The fifth topic was represented by a database for uploading theses presented by students at the final seminar. The final test consisted of 15 questions (5 ones from each basic topic. It was necessary to answer correctly at least 10 questions to receive a certificate of completing the course. Results: The course was put into operation during the summer term of the academic year 2012/2013. To the date 15 of September the total of 23 students enrolled (17, i.e. all students of the elective subject in the Czech version, 2 students of this subject in the English version, 2 postgraduate students and 2 medical doctors. All enrolled students used the course for on-line learning, downloading, or printing course study materials. All undergraduate students were obliged to use it for preparation

  2. Current plans to characterize the design basis ground motion at the Yucca Mountain, Nevada Site

    International Nuclear Information System (INIS)

    Simecka, W.B.; Grant, T.A.; Voegele, M.D.; Cline, K.M.

    1992-01-01

    A site at Yucca Mountain Nevada is currently being studied to assess its suitability as a potential host site for the nation's first commercial high level waste repository. The DOE has proposed a new methodology for determining design-basis ground motions that uses both deterministic and probabilistic methods. The role of the deterministic approach is primary. It provides the level of detail needed by design engineers in the characterization of ground motions. The probabilistic approach provides a logical structured procedure for integrating the range of possible earthquakes that contribute to the ground motion hazard at the site. In addition, probabilistic methods will be used as needed to provide input for the assessment of long-term repository performance. This paper discusses the local tectonic environment, potential seismic sources and their associated displacements and ground motions. It also discusses the approach to assessing the design basis earthquake for the surface and underground facilities, as well as selected examples of the use of this type of information in design activities

  3. Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Katanishi, Shoji; Kunitomi, Kazuhiko

    2003-01-01

    Japan Atomic Energy Research Institute has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is much lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of a depressurization accident. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. (author)

  4. Design Rules for High Temperature Microchemical Systems

    Science.gov (United States)

    2006-10-25

    and heat spreaders , and none were totally satisfactory. Available were either too fragile for use, or required an insulating thickness that was...vacuum pump, but quickly lose reflectivity under the normal conditions. Heat spreaders were also a problem. One would like switchable heat spreaders ...fluid are shown with an amplitude >1400 K. The heat recycle from the hot walls causes a temperature rise much above the adiabatic flame temperature

  5. A proposal to develop a high temperature structural design guideline for HTGR components

    International Nuclear Information System (INIS)

    Hada, K.

    1989-01-01

    This paper presents some proposals for developing a high-temperature structural design guideline for HTGR structural components. It is appropriate that a basis for developing high-temperature structural design rules is rested on well-established elevated-temperature design guidelines, if the same failure modes are expected for high-temperature components as considered in such design guidelines. As for the applicability of ASME B and PV Code Case N-47 to structural design rules for high-temperature components (service temperatures ≥ 900 deg. C), the following critical issues on material properties and service life evaluation rules have been pointed out. (i) no work-hardening of stress-strain curves at high temperatures due to dynamic recrystallization; (ii) issues relating to very significant creep; (iii) ductility loss after long-term ageing at high temperatures; (iv) validity of life-fraction rule (Robinson-Taira rule) as creep-fatigue damage evaluation rule. Furthermore, the validity of design margins of elevated-temperature structural design guidelines to high-temperature design rules should be clarified. Solutions and proposals to these issues are presented in this paper. Concerning no work-hardening due to dynamic recrystallization, it is shown that viscous effects cannot be neglected even at high extension rate for tensile tests, and that changes in viscous deformation rates by dynamic recrystallization should be taken into account. The extension rate for tensile tests is proposed to change at high temperatures. The solutions and proposals to the above-mentioned issues lead to the conclusion that the design methodologies of N-47 are basically applicable to the high-temperature structural design guideline for HTGR structural components in service at about 900 deg. C. (author). 9 refs, 5 figs

  6. Grid fault and design-basis for wind turbines - Final report

    DEFF Research Database (Denmark)

    Hansen, Anca Daniela; Cutululis, Nicolaos Antonio; Markou, Helen

    , have been performed and compared for two cases, i.e. one when the turbine is immediately disconnected from the grid when a grid fault occurs and one when the turbine is equipped with a fault ride-through controller and therefore it is able to remain connected to the grid during the grid fault......This is the final report of a Danish research project “Grid fault and design-basis for wind turbines”. The objective of this project has been to assess and analyze the consequences of the new grid connection requirements for the fatigue and ultimate structural loads of wind turbines....... The fulfillment of the grid connection requirements poses challenges for the design of both the electrical system and the mechanical structure of wind turbines. The development of wind turbine models and novel control strategies to fulfill the TSO’s requirements are of vital importance in this design. Dynamic...

  7. Breckinridge Project, initial effort. Report XI, Volume V. Critical review of the design basis. [Critical review

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-01-01

    Report XI, Technical Audit, is a compendium of research material used during the Initial Effort in making engineering comparisons and decisions. Volumes 4 and 5 of Report XI present those studies which provide a Critical Review of the Design Basis. The Critical Review Report, prepared by Intercontinental Econergy Associates, Inc., summarizes findings from an extensive review of the data base for the H-Coal process design. Volume 4 presents this review and assessment, and includes supporting material; specifically, Design Data Tabulation (Appendix A), Process Flow Sheets (Appendix B), and References (Appendix C). Volume 5 is a continuation of the references of Appendix C. Studies of a proprietary nature are noted and referenced, but are not included in these volumes. They are included in the Limited Access versions of these reports and may be reviewed by properly cleared personnel in the offices of Ashland Synthetic Fuels, Inc.

  8. Summary of the experimental multi-purpose very high temperature gas cooled reactor design

    International Nuclear Information System (INIS)

    1984-12-01

    The report presents the design of Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR) based on the second stage of detailed design which was completed on March 1984, in the from of ''An application of reactor construction permit Appendix 8''. The Experimental VHTR is designed to satisfy with the design specification for the reactor thermal output 50 MW and reactor outlet temperature 950 0 C. The adequacy of the design is also checked by the safety analysis. The planning of plant system and safety is summarized such as safety design requirements and conformance with them, seismic design and plant arrangement. Concerning with the system of the Experimental VHTR the design basis, design data and components are described in the order. (author)

  9. Working Towards Unified Safety Design Criteria for Modular High Temperature Gas-cooled Reactor Designs

    International Nuclear Information System (INIS)

    Reitsma, Frederik; Silady, Fred; Kunitomi, Kazuhiko

    2014-01-01

    The Nuclear Power Development Section of the IAEA recently received approval for a Coordinated Research Project (CRP) to investigate and make proposals on modular High Temperature Gas-cooled Reactor (HTGR) Safety design criteria. It is expected that these criteria would consider past experience and existing safety standards in the light of modular HTGR material and design characteristics to propose safety design criteria. It will consider the deterministic and risk-informed safety design standards that apply to the wide spectrum of Off- normal events under development worldwide for existing and planned HTGRs. The CRP would also take into account lessons from the Fukushima Daiichi accident, clarifying the safety approach and safety evaluation criteria for design and beyond design basis events, including those events that can affect multiple reactor modules and/or are dependent on the application proximate to the plant site. (e. g., industrial process steam/heat). The logical flow of criteria is from the fundamental inherent safety characteristics of modular HTGRs and associated expected performance characteristics, to the safety functions required to ensure those characteristics during the wide spectrum of Off-normal events, and finally to specific criteria related to those functions. This is detailed in the paper with specific examples included of how it may be applied. The results of the CRP will be made available to the member states and HTGR community. (author)

  10. Development of Probabilistic Design Basis Earthquake (DBE) Parameters for Moderate and High Hazard Facilities at INEEL

    International Nuclear Information System (INIS)

    Payne, S. M.; Gorman, V. W.; Jensen, S. A.; Nitzel, M. E.; Russell, M. J.; Smith, R. P.

    2000-01-01

    Design Basis Earthquake (DBE) horizontal and vertical response spectra are developed for moderate and high hazard facilities or Performance Categories (PC) 3 and 4, respectively, at the Idaho National Engineering and Environmental Laboratory (INEEL). The probabilistic DBE response spectra will replace the deterministic DBE response spectra currently in the U.S. Department of Energy Idaho Operations Office (DOE-ID) Architectural Engineering Standards that govern seismic design criteria for several facility areas at the INEEL. Probabilistic DBE response spectra are recommended to DOE Naval Reactors for use at the Naval Reactor Facility at INEEL. The site-specific Uniform Hazard Spectra (UHS) developed by URS Greiner Woodward Clyde Federal Services are used as the basis for developing the DBE response spectra. In 1999, the UHS for all INEEL facility areas were recomputed using more appropriate attenuation relationships for the Basin and Range province. The revised UHS have lower ground motions than those produced in the 1996 INEEL site-wide probabilistic ground motion study. The DBE response spectra were developed by incorporating smoothed broadened regions of the peak accelerations, velocities, and displacements defined by the site-specific UHS. Portions of the DBE response spectra were adjusted to ensure conservatism for the structural design process

  11. A basis for standardized seismic design (SSD) for nuclear power plants/critical facilities

    International Nuclear Information System (INIS)

    O'Hara, T.F.; Jacobson, J.P.; Bellini, F.X.

    1991-01-01

    US Nuclear Power Plants (NPP's) are designed, engineered and constructed to stringent standards. Their seismic adequacy is assured by compliance with regulatory standards and demonstrated by both probabilistic risk assessments (PRAs) and seismic margin studies. However, present seismic siting criteria requires improvement. Proposed changes to siting criteria discussed here will provide a predictable licensing process and a stable regulatory environment. Two recent state-of-the-art studies evaluate the seismic design for all eastern US (EUS) NPP'S: a Lawrence Livermore National Labs study (LLNL, 1989) funded by the NRC and similar research by the Electric Power Research Institute (EPRI, 1989) supported by the utilities. Both confirm that Appendix A 10CFR Part 100 has not provided consistent seismic design levels for all sites. Standardized Seismic Design (SSD) uses a probabilistic framework to accommodate alternative deterministic interpretations. It uses seismic hazard input from EPRI or LLNL to produce consistent bases for future seismic design. SSD combines deterministic and probabilistic insights to provide a comprehensive approach for determining a future site's acceptable seismic design basis

  12. Environmental conditions using thermal-hydraulics computer code GOTHIC for beyond design basis external events

    International Nuclear Information System (INIS)

    Pleskunas, R.J.

    2015-01-01

    In response to the Fukushima Dai-ichi beyond design basis accident in March 2011, the Nuclear Regulatory Commission (NRC) issued Order EA-12-049, 'Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies Beyond-Design-Basis-External-Events'. To outline the process to be used by individual licensees to define and implement site-specific diverse and flexible mitigation strategies (FLEX) that reduce the risks associated with beyond design basis conditions, Nuclear Energy Institute document NEI 12-06, 'Diverse and Flexible Coping Strategies (FLEX) Implementation Guide', was issued. A beyond design basis external event (BDBEE) is postulated to cause an Extended Loss of AC Power (ELAP), which will result in a loss of ventilation which has the potential to impact room habitability and equipment operability. During the ELAP, portable FLEX equipment will be used to achieve and maintain safe shutdown, and only a minimal set of instruments and controls will be available. Given these circumstances, analysis is required to determine the environmental conditions in several vital areas of the Nuclear Power Plant. The BDBEE mitigating strategies require certain room environments to be maintained such that they can support the occupancy of personnel and the functionality of equipment located therein, which is required to support the strategies associated with compliance to NRC Order EA-12-049. Three thermal-hydraulic analyses of vital areas during an extended loss of AC power using the GOTHIC computer code will be presented: 1) Safety-related pump and instrument room transient analysis; 2) Control Room transient analysis; and 3) Auxiliary/Control Building transient analysis. GOTHIC (Generation of Thermal-Hydraulic Information for Containment) is a general purpose thermal-hydraulics software package for the analysis of nuclear power plant containments, confinement buildings, and system components. It is a volume/path/heat sink

  13. A cliff edge evaluation for CANDU-6 beyond design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.M.; Kho, D.W., E-mail: wolsong@khnp.co.kr [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Yi, S.D.; Kang, S.H.; Kim, S.R. [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

    2015-07-01

    The condition of nuclear power plant in the event of station black out (SBO) accompanying large-scale natural disaster exceeding design basis accident (DBA) was evaluated. Additional scenarios were added to the evaluation to review capability of the plant to endure different conditions with different actions. The analysis resulted that the key action required from the operator was to ensure the opening of main steam safety valves (MSSVs) in the secondary side and of motor-operated valves for high pressure injection of Emergency Core Cooling System (HPECCS) to mitigate accidents or extend the cliff edge. (author)

  14. Probabilistic evaluation of concrete containment capacity for beyond design basis internal pressures

    International Nuclear Information System (INIS)

    Tang, H.T.; Dameron, R.A.; Rashid, Y.R.

    1995-01-01

    For beyond design basis internal pressure loading, experimental studies have demonstrated that the most probable failure mode governing the ultimate functional capacity of concrete containments is leak rather than break. Based on leak rates measured in experiments, a prediction formula for leak rate as functions of containment liner size and internal pressure has been postulated. The determination of liner tear is cast in a probabilistic framework. In calculating leakage, particular attention is paid to the evaluation of leakage versus rupture and the loading rates that may be required to leapfrog over a leakage mode. (orig.)

  15. Tests of qualification of national components of nuclear power plants under design basis accident

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1990-01-01

    With the purpose of qualifying national components of nuclear power plants, whose working must be maintained during and after an accident, the Thermohydraulic Division of CDTN have done tests to check the equipment stability, under Design Basis Accident conditions. Until this moment, the following components were tested: electrical junction boxes (connectors); coating systems for wall, inside cover and steel containment; hydraulics components of personnel and equipment airlock. This work describes the test instalation, the tests performed and its results. The components tested, in a general way, fulfil the specified requirements. (author) [pt

  16. Structural analysis for elevated temperature design of the LMFBR

    International Nuclear Information System (INIS)

    Griffin, D.S.

    1976-02-01

    In the structural design of LMFBR components for elevated temperature service it is necessary to take account of the time-dependent, creep behavior of materials. The accommodation of creep to assure design reliability has required (1) development of new design limits and criteria, (2) development of more detailed representations of material behavior, and (3) application of the most advanced analysis techniques. These developments are summarized and examples are given to illustrate the current state of technology in elevated temperature design

  17. Recent advances in design procedures for high temperature plant

    International Nuclear Information System (INIS)

    1988-01-01

    Thirteen papers cover several aspects of design for high temperature plant. These include design codes, computerized structural analysis and mechanical properties of materials at high temperatures. Seven papers are relevant for fast reactors and these are indexed separately. These cover shakedown design, design codes for thin shells subjected to cyclic thermal loading, the inelastic behaviour of stainless steels and creep and crack propagation in reactor structures under stresses caused by thermal cycling loading. (author)

  18. Unites States position paper on sodium fires. Design basis and testing

    International Nuclear Information System (INIS)

    Lancet, R.T.; Johnson, R.P.; Matlin, E.; Vaughan, E.U.; Fields, D.E.; Glueckler, E.; McCormack, J.D.; Miller, C.W.; Pedersen, D.R.

    1989-01-01

    This paper focuses on designs, analyses, and tests performed since the last Sodium Fires Meeting of the IAEA International Working Group on Fast Reactors in May 1982. Since the U.S. Liquid Metal Reactor (LMR) program is focused on the two advanced LMRs, SAFR and PRISM, the paper relates this work to these designs. First, the design philosophy and approach taken by these advanced pool reactors are described. This includes methods of leak detection, the design basis leaks, and passive accommodation of sodium fires. Then the small- and large-scale sodium fire tests performed in support of the Clinch River Breeder Reactor Plant (CRBRP) program, including post-accident cleanup, are presented and related to the advanced LMR designs. Next, the assessment and behavior of the aerosols generated are discussed including generation rate, behavior within structures, release and dispersal, and deposition on safety-grade equipment. Finally, the impact of these aerosols on the performance of safety-grade decay heat removal heat exchange surfaces is discussed including some test results as well as planned tests. (author)

  19. Designing Raster Cells as the Basis for Developing Personal Graphic Language

    Directory of Open Access Journals (Sweden)

    Jana Z. Vujić

    2011-05-01

    Full Text Available Continuous work in creating new designer solutions points towards the need to create personal routines as personalcommunication in the relation comprising design, algorithms, and original computer graphics. This paper showsprocedures for developing a control language for creating graphic designs with individual raster elements (screeningelement obtaint by halftoning. Personal commands should set routines in a language understood by the printer andthe designer. The PostScript basis is used because we mix vector and pixel graphics in the same program stream, aswell as different colour systems, and our own raster forms. The printing raster is set with the target of special designmulti-use, and this includes the field of security graphics and art computer reproduction. Each raster form assumesmodifications, creating their raster family. The raster cell content is transformed with PostScript, allowing the settingof basic values, angle and liniature for each pixel separately. Raster cells are mixed in multi-colour graphics to thelevel of individual designs with variable values of parameters determining them.

  20. Why there is a need to revise the Design Basis Threat concept

    International Nuclear Information System (INIS)

    Kondratov, S.; Steinhausler, F.

    2006-01-01

    The terrorist attacks in the USA on 11 September 2001 necessitated a review of the proven concept of the Design Basis Threat (DBT) for nuclear installations. It can be assumed that revised and upgraded DBT will result in costly technical solutions. Since infrastructure deficits and financial limitations in many countries have already limited the practical application of the DBT, the revised threat assessment is likely to worsen the current unsatisfactory situation. Therefore, a new realism in the use of the DBT concept is proposed based on a three-level approach. This will enable countries to tailor the design of their physical protection systems in accordance with their means by implementing either a minimum required security level protecting only against the most probable threat, or an intermediate protection level reflecting the newly introduced AHARA (As High As Reasonably Achievable) principle, or the optimum protection level based on an externally reviewed, fully comprehensive DBT. (author)

  1. DESIGNING ALGORITHMS FOR SOLVING PHYSICS PROBLEMS ON THE BASIS OF MIVAR APPROACH

    Directory of Open Access Journals (Sweden)

    Dmitry Alekseevich Chuvikov

    2017-05-01

    Full Text Available The paper considers the process of designing algorithms for solving physics problems on the basis of mivar approach. The work also describes general principles of mivar theory. The concepts of parameter, relation and class in mivar space are considered. There are descriptions of properties which every object in Wi!Mi model should have. An experiment in testing capabilities of the Wi!Mi software has been carried out, thus the model has been designed which solves physics problems from year 8 school course in Russia. To conduct the experiment a new version of Wi!Mi 2.1 software has been used. The physics model deals with the following areas: thermal phenomena, electric and electromagnetic phenomena, optical phenomena.

  2. Screening and analysis of beyond design basis accident of 49-2 SPR

    International Nuclear Information System (INIS)

    Zhang Yadong; Guo Yue; Wu Yuanyuan; Zou Yao

    2015-01-01

    The beyond design basis accident was analyzed to ensure safe operation of 49-2 Swimming Pool Reactor (SPR) after design life. Because it's difficult to use PSA method, the unconditional assumed severe accidents were adopted to obtain a conservative result. The main conclusions were obtained by analyzing anticipated transients without scram in station blackout (SBO ATWS), horizontal channel rupture, core uncovering after shutdown and emergency response capacity. The results show that the core is safe in SBO ATWS, and the fuel elements will not melt as long as the core are not exposed in 2.5 h in loss of coolant accident caused by horizontal channel rupture and other factors. The passive siphon breaker function and various ways of emergency core makeup can ensure that the core is not exposed. (authors)

  3. Sea Surface Temperature Modeling using Radial Basis Function Networks With a Dynamically Weighted Particle Filter

    KAUST Repository

    Ryu, Duchwan; Liang, Faming; Mallick, Bani K.

    2013-01-01

    be modeled by a dynamic system which changes with time and location. In this article, we propose a radial basis function network-based dynamic model which is able to catch the nonlinearity of the data and propose to use the dynamically weighted particle

  4. Design of turning hydraulic engines for manipulators of mobile machines on the basis of multicriterial optimization

    Directory of Open Access Journals (Sweden)

    Lagerev I.A.

    2016-12-01

    Full Text Available In this paper the mathematical models of the main types of turning hydraulic engines, which at the present time widely used in the construction of handling systems of domestic and foreign mobile transport-technological machines wide functionality. They allow to take into consideration the most significant from the viewpoint of ensuring high technical-economic indicators of hydraulic efficiency criteria – minimum mass (weight, their volume and losses of power. On the basis of these mathematical models the problem of multicriterial constrained optimization of the constructive sizes of turning hydraulic engines are subject to complex constructive, strength and deformation limits. It allows you to de-velop the hydraulic engines in an optimized design which is required for the purpose of designing a comprehensive measure takes into account efficiency criteria. The multicriterial optimization problem is universal in nature, so when designing a turning hydraulic engines allows for one-, two - and three-criteria optimization without making any changes in the solution algorithm. This is a significant advantage for the development of universal software for the automation of design of mobile transport-technological machines.

  5. Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Katanishi, Shoji; Kunitomi, Kazuhiko

    2003-01-01

    Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300 based on the experience of the High Temperature Engineering Test Reactor (HTTR) of JAERI which is the first High Temperature Gas-cooled Reactor (HTGR) in Japan. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident induced by a large pipe break is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of the depressurization accident. The safety design philosophies for passive cooling system, reactor shutdown system, and so on were determined. The methodology for the safety evaluation, such as safety criteria and selection of events to be evaluated by using estimation of probability of occurrence, were also discussed and determined. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  6. Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2015-11-01

    The demands on nuclear fuel have recently been increasing, and include transient regimes, higher discharge burnup and longer fuel cycles. This has resulted in an increase of loads on fuel and core internals. In order to satisfy these demands while ensuring compliance with safety criteria, new national and international programmes have been launched and advanced modelling codes are being developed. The Fukushima Daiichi accident has particularly demonstrated the need for adequate analysis of all aspects of fuel performance to prevent a failure and also to predict fuel behaviour were an accident to occur.This publication presents the Proceedings of the Technical Meeting on Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents, which was hosted by the Nuclear Power Institute of China (NPIC) in Chengdu, China, following the recommendation made in 2013 at the IAEA Technical Working Group on Fuel Performance and Technology. This recommendation was in agreement with IAEA mid-term initiatives, linked to the post-Fukushima IAEA Nuclear Safety Action Plan, as well as the forthcoming Coordinated Research Project (CRP) on Fuel Modelling in Accident Conditions. At the technical meeting in Chengdu, major areas and physical phenomena, as well as types of code and experiment to be studied and used in the CRP, were discussed. The technical meeting provided a forum for international experts to review the state of the art of code development for modelling fuel performance of nuclear fuel for water cooled reactors with regard to steady state and transient conditions, and for design basis and early phases of severe accidents, including experimental support for code validation. A round table discussion focused on the needs and perspectives on fuel modelling in accident conditions. This meeting was the ninth in a series of IAEA meetings, which reflects Member States’ continuing interest in nuclear fuel issues. The previous meetings were held in 1980 (jointly with

  7. Basis of the tubesheet heat exchanger design rules used in the French pressure vessel code

    International Nuclear Information System (INIS)

    Osweiller, F.

    1990-01-01

    For about 40 years most tubesheet heat exchangers have been designed according to the standards of TEMA. Partly due to their simplicity, these rules do not assure a safe heat-exchangers design in all cases. This is the main reason why new tubesheet design rules were developed in 1981 in France for the French pressure vessel code CODAP. For fixed tubesheet heat exchangers the new rules account for the elastic rotational restraint of the shell and channel at the outer edge of the tubesheet. For floating-head and U- tube exchangers an approach was selected with some modifications. In both cases the tubesheet is replaced by an equivalent solid plate with adequate effective elastic constants, and the tube bundle is simulated by an elastic foundation. The elastic restraint at the edge of the tubesheet due the shell and channel is accounted for in different ways in the two types of heat exchangers. The purpose of the paper is to present the main basis of these rules and to compare them to TEMA rules

  8. Technical meeting on progress in managing, and limiting the consequences of events exceeding the design basis

    International Nuclear Information System (INIS)

    Fabian, H.

    2004-01-01

    The Technical Groups on 'Reactor Safety' and 'Thermodynamics and Fluid Dynamics' of the Kerntechnische Gesellschaft e.V. organized a joint technical meeting on 'Progress in Managing, and Limiting the Consequences of, Events Exceeding the Design Basis' at the FTU Training Center of the Karlsruhe Research Center. The topic chosen, the papers presented, the presenters, and the non-technical part of the program met with lively interest on the part of institutions in the nuclear field. These were the objectives of the technical meeting: - Establishing a forum for communicating relevant topics. - In-depth discussion of the main topic, i.e. the advanced development of reactor safety, research in the field, and its application, in twenty selected papers presented by speakers from different institutions. - Presentation of topical work in a nuclear technology institution, the Karlsruhe Research Center. (orig.) [de

  9. Toxic industrial chemicals (TICs) as asymmetric weapons: the design basis threat

    International Nuclear Information System (INIS)

    Skinner, L.

    2009-01-01

    Asymmetric warfare concepts relate well to the use of improvised chemical weapons against urban targets. Sources of information on toxic industrial chemicals (TICs) and lists of high threat chemicals are available that point to likely choices for an attack. Accident investigations can be used as a template for attacks, and to judge the possible effectiveness of an attack using TICs. The results of a chlorine rail car accident in South Carolina, USA and the Russian military assault on a Moscow theater provide many illustrative points for similar incidents that mighty be carried out deliberately. Computer modeling of outdoor releases shows how an attack might take into consideration issues of stand-off distance and dilution. Finally, the preceding may be used to estimate with some accuracy the design basis threat posed by the used of TICs as weapons.(author)

  10. Facts learnt from the Hanshin-Awaji disaster and consideration on design basis earthquake

    International Nuclear Information System (INIS)

    Shibata, Heki

    1997-01-01

    This paper will deal with how to establish the concept of the design basis earthquake for critical industrial facilities such as nuclear power plants in consideration of disasters induced by the 1995 Hyogoken-Nanbu Earthquake (Southern Hyogo-prefecture Earthquake-1995), so-called Kobe earthquake. The author once discussed various DBEs at 7 WCEE. At that time, the author assumed that the strongest effective PGA would be 0.7 G, and compared to the values of accelerations to a structure obtained by various codes in Japan and other countries. The maximum PGA observed by an instrument at the Southern Hyogo-pref. Earthquake-1995 exceeded the previous assumption of the author, even though the evaluation results of the previous paper had been pessimistic. According to the experience of Kobe event, the author will point out the necessity of the third earthquake S s adding to S 1 and S 2 , previous DBEs. (author)

  11. Facts learnt from the Hanshin-Awaji disaster and consideration on design basis earthquake

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Heki [Yokohama National Univ. (Japan). Faculty of Engineering

    1997-03-01

    This paper will deal with how to establish the concept of the design basis earthquake for critical industrial facilities such as nuclear power plants in consideration of disasters induced by the 1995 Hyogoken-Nanbu Earthquake (Southern Hyogo-prefecture Earthquake-1995), so-called Kobe earthquake. The author once discussed various DBEs at 7 WCEE. At that time, the author assumed that the strongest effective PGA would be 0.7 G, and compared to the values of accelerations to a structure obtained by various codes in Japan and other countries. The maximum PGA observed by an instrument at the Southern Hyogo-pref. Earthquake-1995 exceeded the previous assumption of the author, even though the evaluation results of the previous paper had been pessimistic. According to the experience of Kobe event, the author will point out the necessity of the third earthquake S{sub s} adding to S{sub 1} and S{sub 2}, previous DBEs. (author)

  12. An overview of the UK regulatory expectation for design basis accident analysis

    International Nuclear Information System (INIS)

    Trimble, Andy

    2013-01-01

    The UK Health and Safety Executive published its most recent regulatory expectations in the 2006 version of it's safety assessment principles (SAPs). This built on experience regulating the full range of facilities for which it is responsible. Thus the principles underpinning all regulatory safety case assessment are the same but the implementation differs depending on the application. This paper will describe the published design basis accident analysis (DBAA) logic in context with other technical aspects of the regulatory expectation for safety cases. It will further illustrate the DBAA methodology with practical examples from actual experience on reprocessing plant gained over the last 15 years or so. Among the examples will be the relevance of conventional safety fault initiators to nuclear safety assessment. It will further demonstrate the derivation of facility limits and conditions necessary for nuclear safety. (authors)

  13. Design Basis Knowledge Management for New Build Projects & Ageing Plants - A Perspective

    International Nuclear Information System (INIS)

    Weightman, Mike

    2013-01-01

    Summary: • KM for Design Basis of New and Ageing nuclear facilities is at a crossroads; • Needs leadership, vision, cultural change and resources; • Outcome of this workshop is vital; • Information is not knowledge; • Knowledge includes the WHAT, the HOW, the WHY, the Environment and, importantly, Application; • In general, Industry and Regulators are behind the curve; • Develop and apply the principles rigorously; • Keep it simple - focus first on Leadership, values (e.g. questioning attitude), culture, and prioritise – risk informed; • KM is a complex organic creature and needs to be nurtured, fed, learn, grow, evolve in response to a changing environment, and discharge what is not needed to prosper

  14. Maximal design basis accident of fusion neutron source DEMO-TIN

    Energy Technology Data Exchange (ETDEWEB)

    Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

  15. Beyond Design Basis Severe Accident Management as an Element of DiD Concept Strengthening

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsov, M., E-mail: kuznetsov_mv@vosafety.ru [FSUE VO “Safety”, Moscow (Russian Federation)

    2014-10-15

    The 4{sup th} Level of DiD is ensured by management of beyond design basis accidents which is achieved by implementation of the Beyond Design Basis Accidents Management Guidance (BDBAMG) and, if necessary, by additional technical devices and organizational measures at NPP Unit. BDBAMG is located between Levels 3 and 5 in DiD and is related to them. It is connected with Level 3 by means of conditions generated at this Level and according to which BDBAM should be initiated (Level 4). It is associated with Level 5 by conditions which necessitate implementation of Emergency planning. Both types of conditions should be identified in BDBAMG. BDBAs including the phase of severe damage of fuel and protective barriers (severe accidents) in accordance with Russian regulatory framework are a subset of all BDBAs set. In this connection, such accident scenarios meet the representativeness criterion for further analysis and development of Guidance for their management. BDBAMG availability, as it provides robustness of DiD as a whole, is an obligatory condition for obtaining a NPP operational license. In the process of BDBAMG development and implementation a feedback with technical and organizational measures, comprising Level 1 and, to a less extent, Level 2, comes up. BDBAMG verification is an important final stage of its development. Addressing severe accidents, it is a challenging issue for a full scope simulator and may require its software modernization to make it responsive to severe accident phenomena. The existing BDBAMGs should be updated due to NPP Unit modernizations and in conjunction with the latest knowledge on severe accident phenomenology and lessons learnt from known events (e.g. NPP Fukushima). Thus, improvements incorporated in BDBAMG, enhance the strength of DiD. (author)

  16. Overview of Mobile Equipment Used in Case of Beyond Design Basis Accident at NPP Krsko

    International Nuclear Information System (INIS)

    Lukacevic, H.; Kopinc, D.; Ivanjko, M.

    2016-01-01

    Terrorist attack in USA in the September 11, 2001 and accident at the Fukushima - Daiichi Nuclear Power Station in the March 11, 2011 highlight the importance of mitigating strategies in responding to Beyond Design Basis Accident (BDBA), while ensuring cooling of reactor core, containment and spent fuel pool. Nuclear Power Plant Krsko (NEK) has acquired additional mobile equipment and made necessary modifications on existing systems for the connection of this equipment (fast couplers). Usage of mobile equipment is not only limited to design basis accident (DBA), but, also to prevent and mitigate the consequences in case of BDBA, when other plant systems are not available. NEK also decided to take steps for upgrade of safety measures and prepared Safety Upgrade Program (SUP), which is consistent with the nuclear industry response to the Fukushima accident and is implementing main projects and modifications related to SUP. NEK mobile equipment is not required to operate under normal reactor plant operation except for periodic surveillance testing and is incorporated into the normal training process. Equipment is dislocated from the reactor building and most of the equipment is located in the new building, able to withstand extreme natural events, including earthquakes and tornadoes. The usage of all mobile equipment is prescribed as an additional option in NEK operating procedures in following cases and enables following options: filling various tanks, filling the steam generators, filling the containment, additional compressed air source, spent fuel pool refilling and spraying, alternative power supply. This document provides an overview of NEK mobile equipment, which consists of various mobile fire protection pumps, air compressors, protective equipment, fire trucks, diesel generators. Sufficient fuel supply for the equipment is provided on site for a minimum three days of operation. (author).

  17. Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors

    International Nuclear Information System (INIS)

    Chvetsov, I.; Kouznetsov, I.; Volkov, A.

    2000-01-01

    GRIF-SM code was developed at the IPPE fast reactor department in 1992 for the analysis of transients in sodium cooled fast reactors under severe accident conditions. This code provides solution of transient hydrodynamics and heat transfer equations taking into account possibility of coolant boiling, fuel and steel melting, reactor kinetics and reactivity feedback due to variations of the core components temperature, density and dimensions. As a result of calculation, transient distribution of the coolant velocity and density was determined as well as temperatures of the fuel pins, reactor core and primary circuit as a whole. Development of the code during further 6 years period was aimed at the modification of the models describing thermal hydraulic characteristics of the reactor, and in particular in detailed description of the sodium boiling process. The GRIF-SM code was carefully validated against FZK experimental data on steady state sodium boiling in the electrically heated tube; transient sodium boiling in the 7-pin bundle; transient sodium boiling in the 37-pin bundle under flow redaction simulating ULOF accident. To show the code capabilities some results of code application for beyond design basis accident analysis on BN-800-type reactor are presented. (author)

  18. Design of temperature monitoring system based on CAN bus

    Science.gov (United States)

    Zhang, Li

    2017-10-01

    The remote temperature monitoring system based on the Controller Area Network (CAN) bus is designed to collect the multi-node remote temperature. By using the STM32F103 as main controller and multiple DS18B20s as temperature sensors, the system achieves a master-slave node data acquisition and transmission based on the CAN bus protocol. And making use of the serial port communication technology to communicate with the host computer, the system achieves the function of remote temperature storage, historical data show and the temperature waveform display.

  19. Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

    International Nuclear Information System (INIS)

    Wei, B.C.; Cooper, W.L. Jr.; Dhalla, A.K.

    1987-09-01

    Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work

  20. Application of the MELCOR code to design basis PWR large dry containment analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  1. Why there is a need to revise the design basis threat concept

    International Nuclear Information System (INIS)

    Kondratov, S.; Steinhaeusler, F.

    2005-01-01

    Full text: The coordinated terrorist attacks in the United States on September 11, 2001, necessitated the review of the proven concept of the Design Basis Threat (DBT) for nuclear installations. It is safe to assume that revised and upgraded DBT will result in costly technical solutions. Since infrastructure deficits and financial limitations in many countries have already limited the practical application of the DBT in many instances, the revised threat assessment is likely to worsen the current dissatisfactory situation further. Therefore, a new realism in the use of the DBT concept is proposed, based on a three-level approach. This will enable countries to tailor the design of their physical protection systems in accordance with their means by implementing either a minimum required security level protecting only against the most probable threat, or aiming for an intermediate protection level reflecting the newly introduced AHARA - as high as reasonably achievable - principle, or providing the optimum protection level based on an externally reviewed, fully comprehensive DBT. (author)

  2. Comparison of deterministic and stochastic techniques for estimation of design basis floods for nuclear power plants

    International Nuclear Information System (INIS)

    Solomon, S.I.; Harvey, K.D.

    1982-12-01

    The IAEA Safety Guide 50-SG-S10A recommends that design basis floods be estimated by deterministic techniques using probable maximum precipitation and a rainfall runoff model to evaluate the corresponding flood. The Guide indicates that stochastic techniques are also acceptable in which case floods of very low probability have to be estimated. The paper compares the results of applying the two techniques in two river basins at a number of locations and concludes that the uncertainty of the results of both techniques is of the same order of magnitude. However, the use of the unit hydrograph as the rainfall runoff model may lead in some cases to nonconservative estimates. A distributed non-linear rainfall runoff model leads to estimates of probable maximum flood flows which are very close to values of flows having a 10 6 - 10 7 years return interval estimated using a conservative and relatively simple stochastic technique. Recommendations on the practical application of Safety Guide 50-SG-10A are made and the extension of the stochastic technique to ungauged sites and other design parameters is discussed

  3. Comparison of deterministic and stochastic techniques for estimation of design basis floods for nuclear power plants

    International Nuclear Information System (INIS)

    Solomon, S.I.; Harvey, K.D.; Asmis, G.J.K.

    1983-01-01

    The IAEA Safety Guide 50-SG-S10A recommends that design basis floods be estimated by deterministic techniques using probable maximum precipitation and a rainfall runoff model to evaluate the corresponding flood. The Guide indicates that stochastic techniques are also acceptable in which case floods of very low probability have to be estimated. The paper compares the results of applying the two techniques in two river basins at a number of locations and concludes that the uncertainty of the results of both techniques is of the same order of magnitude. However, the use of the unit hydrograph as the rain fall runoff model may lead in some cases to non-conservative estimates. A distributed non-linear rainfall runoff model leads to estimates of probable maximum flood flows which are very close to values of flows having a 10 6 to 10 7 years return interval estimated using a conservative and relatively simple stochastic technique. Recommendations on the practical application of Safety Guide 50-SG-10A are made and the extension of the stochastic technique to ungauged sites and other design parameters is discussed

  4. Considerations on Fail Safe Design for Design Basis Accident (DBA) vs. Design Extension Condition (DEC): Lesson Learnt from the Fukushima Accident

    International Nuclear Information System (INIS)

    Ha, Jun Su; Kim, Sungyeop

    2014-01-01

    The fail safety design is referred to as an inherently safe design concept where the failure of an SSC (System, Structure or Component) leads directly to a safe condition. Usually the fail safe design has been devised based on the design basis accident (DBAs), because the nuclear safety has been assured by securing the capability to safely cope with DBAs. Currently regards have been paid to the DEC (Design Extension Condition) as an extended design consideration. Hence additional attention should be paid to the concept of the fail safe design in order to consider the DEC, accordingly. In this study, a case chosen from the Fukushima accident is studied to discuss the issue associated with the fail safe design in terms of DBA and DEC standpoints. For the fail safe design to be based both on the DBA and the DEC, a Mode Changeable Fail Safe Design (MCFSD) is proposed in this study. Additional discussions on what is needed for the MCFSD to be applied in the nuclear safety are addressed as well. One of the lessons learnt from the Fukushima accident should include considerations on the fail-safe design in a changing regulatory framework. Currently the design extension condition (DEC) including severe accidents should be considered during designing and licensing NPPs. Hence concepts on the fail safe design need to be changed to be based on not only the DBA but also the DEC. In this study, a case on a fail-safe design chosen from the Fukushima accident is studied to discuss the issue associated with the fail safe design in terms of DBA and DEC conditions. For the fail safe design to be based both on the DBA and the DEC, a Mode Changeable Fail Safe Design (MCFSD) is proposed in this study. Additional discussions on what is needed for the MCFSD to be applied in the nuclear safety are addressed as well

  5. Design and development of gas turbine high temperature reactor 300

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Takizuka, Takakazu

    2003-01-01

    JAERI (Japan Atomic Energy Research Institute) has been designing a Japan's original gas turbine high temperature reactor, GTHTR300 (Gas Turbine High Temperature Reactor 300). The greatly simplified design based on salient features of the HTGR (High Temperature Gas-cooled reactor) with a closed helium gas turbine enables the GTHTR300 a high efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the HTTR (High Temperature Engineering Test Reactor) and fossil gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original features of this system are core design with two-year refueling interval, conventional steel material usage for a reactor pressure vessel, innovative plant flow scheme and horizontally installed gas turbine unit. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200 thousands Yen/kWe, and the electric generation cost is close to a target cost of 4 Yen/kWh. This paper describes the original design features focusing on reactor core design, fuel design, in-core structure design and reactor pressure vessel design except PCU design. Also, R and D for developing the power conversion unit is briefly described. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  6. Design and utilisation of protocols to characterise dynamic PET uptake of two tracers using basis pursuit

    Science.gov (United States)

    Bell, Christopher; Puttick, Simon; Rose, Stephen; Smith, Jye; Thomas, Paul; Dowson, Nicholas

    2017-06-01

    Imaging using more than one biological process using PET could be of great utility, but despite previously proposed approaches to dual-tracer imaging, it is seldom performed. The alternative of performing multiple scans is often infeasible for clinical practice or even in research studies. Dual-tracer PET scanning allows for multiple PET radiotracers to be imaged within the same imaging session. In this paper we describe our approach to utilise the basis pursuit method to aid in the design of dual-tracer PET imaging experiments, and later in separation of the signals. The advantage of this approach is that it does not require a compartment model architecture to be specified or even that both signals are distinguishable in all cases. This means the method for separating dual-tracer signals can be used for many feasible and useful combinations of biology or radiotracer, once an appropriate scanning protocol has been decided upon. Following a demonstration in separating the signals from two consecutively injected radionuclides in a controlled experiment, phantom and list-mode mouse experiments demonstrated the ability to test the feasibility of dual-tracer imaging protocols for multiple injection delays. Increases in variances predicted for kinetic macro-parameters V D and K I in brain and tumoral tissue were obtained when separating the synthetically combined data. These experiments confirmed previous work using other approaches that injections delays of 10-20 min ensured increases in variance were kept minimal for the test tracers used. On this basis, an actual dual-tracer experiment using a 20 min delay was performed using these radio tracers, with the kinetic parameters (V D and K I) extracted for each tracer in agreement with the literature. This study supports previous work that dual-tracer PET imaging can be accomplished provided certain constraints are adhered to. The utilisation of basis pursuit techniques, with its removed need to specify a model

  7. Thermocouple design for measuring temperatures of small insects

    Science.gov (United States)

    A.A. Hanson; R.C. Venette

    2013-01-01

    Contact thermocouples often are used to measure surface body temperature changes of insects during cold exposure. However, small temperature changes of minute insects can be difficult to detect, particularly during the measurement of supercooling points. We developed two thermocouple designs, which use 0.51 mm diameter or 0.127 mm diameter copper-constantan wires, to...

  8. Design of High Temperature Reactor Vessel Using ANSYS Software

    International Nuclear Information System (INIS)

    Bandriyana; Kasmudin

    2003-01-01

    Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600 o C and the outer temperature of 500 and 600 o C. Result of calculation gave the maximum stress for outer temperature of 600 o C was 288 N/ mm 2 and strain of 0.000187. For outer temperature of 500 o C the maximum stress was 576 N/ mm 2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm 2 can be used for vessel material with outer wall temperature of 600 o C

  9. Development, Use and Maintenance of the Design Basis Threat. Implementing Guide (Arabic Edition)

    International Nuclear Information System (INIS)

    2009-01-01

    threat to those assets. As described in this publication, an understanding of the threat can lead to a detailed description of potential adversaries (the design basis threat), which, in turn, is the basis of an appropriately designed physical protection system. This direct link gives confidence that protection would be effective against an adversary attack. International experience in using a design basis threat to protect assets of high consequence is largely based on the protection of nuclear material and facilities. Furthermore, the nuclear security documents defining and recommending that physical protection be based upon the threat. The Physical Protection Objectives and Fundamental Principles, the Recommendations on the Physical Protection of Nuclear Facilities and Nuclear Material, and the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended - do so exclusively for the protection of nuclear material and facilities. Given the historical background, and its continuing contemporary relevance, it has been necessary to draw on that nuclear protection experience in developing this publication. However, the general approach can also be applied to protecting other assets that require a high degree of confidence in the effectiveness of their protection, such as high-activity radioactive material. Specialists from France, Germany, Japan, the Russian Federation, Spain, the United Kingdom, and the United States of America assisted the IAEA in preparing this publication. A draft was presented to an open-ended technical meeting in December 2006, and subsequently circulated for comment to all Member States. This publication is consistent with The Physical Protection Objectives and Fundamental Principles; the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended; and the Recommendations on the Physical Protection of Nuclear Facilities and Nuclear Material.

  10. Development, use and maintenance of the design basis threat. Implementing guide

    International Nuclear Information System (INIS)

    2009-01-01

    threat to those assets. As described in this publication, an understanding of the threat can lead to a detailed description of potential adversaries (the design basis threat), which, in turn, is the basis of an appropriately designed physical protection system. This direct link gives confidence that protection would be effective against an adversary attack. International experience in using a design basis threat to protect assets of high consequence is largely based on the protection of nuclear material and facilities. Furthermore, the nuclear security documents defining and recommending that physical protection be based upon the threat - The Physical Protection Objectives and Fundamental Principles (GOV/2001/41/ Attachment), the Recommendations on the Physical Protection of Nuclear Facilities and Nuclear Material (INFCIRC/225/Rev. 4 (corrected)), and the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended (INFCIRC/274) (adopted on 8 July 2005; (GOV/2005/57)) - do so exclusively for the protection of nuclear material and facilities. Given the historical background, and its continuing contemporary relevance, it has been necessary to draw on that nuclear protection experience in developing this publication. However, the general approach can also be applied to protecting other assets that require a high degree of confidence in the effectiveness of their protection, such as high-activity radioactive material. Specialists from France, Germany, Japan, the Russian Federation, Spain, the United Kingdom, and the United States of America assisted the IAEA in preparing this publication. A draft was presented to an open-ended technical meeting in December 2006, and subsequently circulated for comment to all Member States. This publication is consistent with The Physical Protection Objectives and Fundamental Principles; the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended; and the Recommendations on the

  11. Design basis and design features of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia)

    International Nuclear Information System (INIS)

    1994-05-01

    The prime objective of the IAEA Technical Co-operation Project on Evaluation of Safety Aspects of WWER-440 model 213 NPPs is to co-ordinate and to integrate assistance to national organizations in studying selected aspects of safety for the same type of reactors. Consequently, the study integrated the results generated by national activities carried out in the Czech Republic, Hungary, Slovakia and Ukraine and co-ordinated through the IAEA. Valuable assistance in carrying out the tasks was also provided by Bulgaria and Poland. A set of publications is being prepared to present the results of the project. The publications are intended to facilitate the review and utilization of the results of the project. They are also providing assistance in further refinement and/or extension of plant specific safety evaluation of model 213 NPPs. This Technical Document addressing the design basis and safety related design features of WWER-440 model 213 plants is the first of the series to be published. It is hoped that this document will be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, 36 figs, tabs

  12. SRS BEDROCK PROBABILISTIC SEISMIC HAZARD ANALYSIS (PSHA) DESIGN BASIS JUSTIFICATION (U)

    Energy Technology Data Exchange (ETDEWEB)

    (NOEMAIL), R

    2005-12-14

    This represents an assessment of the available Savannah River Site (SRS) hard-rock probabilistic seismic hazard assessments (PSHAs), including PSHAs recently completed, for incorporation in the SRS seismic hazard update. The prior assessment of the SRS seismic design basis (WSRC, 1997) incorporated the results from two PSHAs that were published in 1988 and 1993. Because of the vintage of these studies, an assessment is necessary to establish the value of these PSHAs considering more recently collected data affecting seismic hazards and the availability of more recent PSHAs. This task is consistent with the Department of Energy (DOE) order, DOE O 420.1B and DOE guidance document DOE G 420.1-2. Following DOE guidance, the National Map Hazard was reviewed and incorporated in this assessment. In addition to the National Map hazard, alternative ground motion attenuation models (GMAMs) are used with the National Map source model to produce alternate hazard assessments for the SRS. These hazard assessments are the basis for the updated hard-rock hazard recommendation made in this report. The development and comparison of hazard based on the National Map models and PSHAs completed using alternate GMAMs provides increased confidence in this hazard recommendation. The alternate GMAMs are the EPRI (2004), USGS (2002) and a regional specific model (Silva et al., 2004). Weights of 0.6, 0.3 and 0.1 are recommended for EPRI (2004), USGS (2002) and Silva et al. (2004) respectively. This weighting gives cluster weights of .39, .29, .15, .17 for the 1-corner, 2-corner, hybrid, and Greens-function models, respectively. This assessment is judged to be conservative as compared to WSRC (1997) and incorporates the range of prevailing expert opinion pertinent to the development of seismic hazard at the SRS. The corresponding SRS hard-rock uniform hazard spectra are greater than the design spectra developed in WSRC (1997) that were based on the LLNL (1993) and EPRI (1988) PSHAs. The

  13. Simulation of the passive UHF devices on the basis of high-temperature superconductors for planar multilayer anisotropic structures

    CERN Document Server

    Gashinova, M S; Kolmakov, Y A; Vendik, I B

    2002-01-01

    The electrodynamic analysis of the arbitrary multilayer medium, including the anisotropic layers and containing the arbitrary form conductors is carried out. Thin layers of the high-temperature superconductor (HTSC) are considered as conductors. Determination of the surface current density is a result of the numerical solution. Accounting for the losses in the HTSC is accomplished on the basis of determining the equivalent surface impedance and using the Leontovich boundary conditions. Anisotropy is accounted for in the determination of the Green spectral dyad for the structure with arbitrary number of the anisotropic or isotropic layers. Calculation of the surface current density distribution demonstrates the correctness of the proposed model

  14. Design for ASIC reliability for low-temperature applications

    Science.gov (United States)

    Chen, Yuan; Mojaradi, Mohammad; Westergard, Lynett; Billman, Curtis; Cozy, Scott; Burke, Gary; Kolawa, Elizabeth

    2005-01-01

    In this paper, we present a methodology to design for reliability for low temperature applications without requiring process improvement. The developed hot carrier aging lifetime projection model takes into account both the transistor substrate current profile and temperature profile to determine the minimum transistor size needed in order to meet reliability requirements. The methodology is applicable for automotive, military, and space applications, where there can be varying temperature ranges. A case study utilizing this methodology is given to design for reliability into a custom application-specific integrated circuit (ASIC) for a Mars exploration mission.

  15. Fundamental basis and implementation of shell and tube heat exchanger project design: condenser and evaporator study

    Science.gov (United States)

    Dalkilic, A. S.; Acikgoz, O.; Tapan, S.; Wongwises, S.

    2016-12-01

    A shell and tube heat exchanger is used as a condenser and an evaporator in this theoretical study. Parametric performance analyses for various actual refrigerants were performed using well-known correlations in open sources. Condensation and evaporation were occurred in the shell side while the water was flowing in the tube side of heat exchanger. Heat transfer rate from tube side was kept constant for condenser and evaporator design. Condensing temperatures were varied from 35 to 60 °C whereas evaporating temperatures were ranging from -15 to 10 °C for the refrigerants of R12, R22, R134a, R32, R507A, R404A, R502, R407C, R152A, R410A and R1234ZE. Variation of convective heat transfer coefficients of refrigerants, total heat transfer coefficients with Reynolds numbers and saturation temperatures were given as validation process considering not only fouling resistance and omission of it but also staggered (triangular) and line (square) arrangements. The minimum tube lengths and necessary pumping powers were calculated and given as case studies for the investigated refrigerants considering validation criteria. It was understood that refrigerant type, fouling resistance and arrangement type are one of the crucial issues regarding the determination of heat exchanger's size and energy consumption. Consequently, R32 and R152a were found to require the shortest tube length and lowest pumping power in the condenser, whereas R507 and R407C have the same advantages in the evaporator. Their heat transfer coefficients were also determined larger than others as expectedly.

  16. Guide line for operator in beyond design basis events for AHWR

    International Nuclear Information System (INIS)

    Kumar, Mithilesh; Mukhopadhyay, D.; Lele, H.G.; Vaze, K.K.

    2011-01-01

    Enhanced defence-in-depth is incorporated in the proposed Advanced Heavy Water Reactor (AHWR) as a part of their fundamental safety approach to ensure that the levels of protection in defence-in-depth shall be more independent from each other than in existing installation. Safety is enhanced by incorporating into their designs, increased emphasis on inherently safe characteristics and passive systems as a part of their fundamental safety approach. It is ensured that the risk from radiation exposures to workers, the public and the environment during construction/commissioning, operation, and decommissioning, shall be comparable to that of other industrial facilities used for similar purposes. This implies that there will be no need for relocation or evacuation measures outside the plant site, apart from those generic emergency measures developed for any industrial facility. It has been demonstrated by analyses that there is no core damage for PIEs with frequencies more than 10- 10 /year. However some scenarios in residual risk domain are considered to demonstrate that dose at plant boundary is within prescribed acceptable limit. It is also possible to arrest core damage progression at various stages of event progression, by incorporating certain operating procedures, without any release. This paper discusses analyses of such low frequency event with multiple failure under the category of 'Decrease in MHT inventory' where plant related symptoms like channel exit temperature, channel component temperatures, moderator level etc. with respect to time are quantified. The operator guide line has been given for case like Loss of coolant without Emergency core coolant system (ECCS) and loss moderator heat sink. It has been observed that 3.0 kg/s mass flow rate is adequate to capture the rising trend of clad surface temperature. (author)

  17. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs

  18. Grid fault and design-basis for wind turbines. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, A.D.; Cutululis, N.A.; Markou, H.; Soerensen, Poul; Iov, F.

    2010-01-15

    This is the final report of a Danish research project 'Grid fault and design-basis for wind turbines'. The objective of this project has been to assess and analyze the consequences of the new grid connection requirements for the fatigue and ultimate structural loads of wind turbines. The fulfillment of the grid connection requirements poses challenges for the design of both the electrical system and the mechanical structure of wind turbines. The development of wind turbine models and novel control strategies to fulfill the TSO's requirements are of vital importance in this design. Dynamic models and different fault ride-through control strategies have been developed and assessed in this project for three different wind turbine concepts (active stall wind turbine, variable speed doublyfed induction generator wind turbine, variable speed multipole permanent magnet wind turbine). A computer approach for the quantification of the wind turbines structural loads caused by the fault ride-through grid requirement, has been proposed and exemplified for the case of an active stall wind turbine. This approach relies on the combination of knowledge from complimentary simulation tools, which have expertise in different specialized design areas for wind turbines. In order to quantify the impact of the grid faults and grid requirements fulfillment on wind turbines structural loads and thus on their lifetime, a rainflow and a statistical analysis for fatigue and ultimate structural loads, respectively, have been performed and compared for two cases, i.e. one when the turbine is immediately disconnected from the grid when a grid fault occurs and one when the turbine is equipped with a fault ride-through controller and therefore it is able to remain connected to the grid during the grid fault. Different storm control strategies, that enable variable speed wind turbines to produce power at wind speeds higher than 25m/s and up to 50m/s without substantially increasing

  19. Ideal flow theory for the double - shearing model as a basis for metal forming design

    Science.gov (United States)

    Alexandrov, S.; Trung, N. T.

    2018-02-01

    In the case of Tresca’ solids (i.e. solids obeying the Tresca yield criterion and its associated flow rule) ideal flows have been defined elsewhere as solenoidal smooth deformations in which an eigenvector field associated everywhere with the greatest principal stress (and strain rate) is fixed in the material. Under such conditions all material elements undergo paths of minimum plastic work, a condition which is often advantageous for metal forming processes. Therefore, the ideal flow theory is used as the basis of a procedure for the preliminary design of such processes. The present paper extends the theory of stationary planar ideal flow to pressure dependent materials obeying the double shearing model and the double slip and rotation model. It is shown that the original problem of plasticity reduces to a purely geometric problem. The corresponding system of equations is hyperbolic. The characteristic relations are integrated in elementary functions. In regions where one family of characteristics is straight, mapping between the principal lines and Cartesian coordinates is determined by linear ordinary differential equations. An illustrative example is provided.

  20. Development of dose calculation program (DBADOSE) incorporating alternative source term due to design basis accident

    International Nuclear Information System (INIS)

    Bae, Young Jig; Nam, Ki Mun; Lee, Yu Jong; Chung, Chan Young

    2003-01-01

    Source terms presented in TID-14844 and Regulatory Guide 1.4 have been used for radiological analysis of design basis accidents for licensing existing pressurized water reactor (PWR). However, more realistic and physically-based source term based on results of study and experiments for about 30 years after the publication of TID-14844 was developed and presented in NUREG-1465 published by U.S NRC in 1995. In addition, ICRP has revised dose concepts and criteria through the publication of ICRP-9, 26, 60 and recommended effective dose concepts rather than critical organ concept since the publication of ICRP-26. Accordingly, multipurpose computer program called DBADOSE incorporating alternative source terms in NUREG-1465 and effective dose concepts in ICRP-60 was developed. Comparison of results of DBADOSE with those of POSTDBA and STARDOSE was performed and verified and no significant difference and inaccuracy were found. DBADOSE will be used to evaluate accidental doses for licensing application according to the domestic laws that are expected to be revised in the near future

  1. iROCS: Integrated accident management framework for coping with beyond-design-basis external events

    International Nuclear Information System (INIS)

    Kim, Jaewhan; Park, Soo-Yong; Ahn, Kwang-Il; Yang, Joon-Eon

    2016-01-01

    Highlights: • An integrated mitigating strategy to cope with extreme external events, iROCS, is proposed. • The strategy aims to preserve the integrity of the reactor vessel as well as core cooling. • A case study for an extreme damage state is performed to assess the effectiveness and feasibility of candidate mitigation strategies under an extreme event. - Abstract: The Fukushima Daiichi accident induced by the Great East Japan earthquake and tsunami on March 11, 2011, poses a new challenge to the nuclear society, especially from an accident management viewpoint. This paper presents a new accident management framework called an integrated, RObust Coping Strategy (iROCS) to cope with beyond-design-basis external events (BDBEEs). The iROCS approach is characterized by classification of various plant damage conditions (PDCs) that might be impacted by BDBEEs and corresponding integrated coping strategies for each of PDCs, aiming to maintain and restore core cooling (i.e., to prevent core damage) and to maintain the integrity of the reactor pressure vessel if it is judged that core damage may not be preventable in view of plant conditions. From a case study for an extreme damage condition, it showed that candidate accident management strategies should be evaluated from the viewpoint of effectiveness and feasibility against accident scenarios and extreme damage conditions of the site, especially when employing mobile or portable equipment under BDBEEs within the limited time available to achieve desired goals such as prevention of core damage as well as a reactor vessel failure.

  2. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Banati, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  3. Design of elliptic curve cryptoprocessors over GF(2^163 using the Gaussian normal basis

    Directory of Open Access Journals (Sweden)

    Paulo Cesar Realpe

    2014-05-01

    Full Text Available This paper presents the efficient hardware implementation of cryptoprocessors that carry out the scalar multiplication kP over finite field GF(2163 using two digit-level multipliers. The finite field arithmetic operations were implemented using Gaussian normal basis (GNB representation, and the scalar multiplication kP was implemented using Lopez-Dahab algorithm, 2-NAF halve-and-add algorithm and w-tNAF method for Koblitz curves. The processors were designed using VHDL description, synthesized on the Stratix-IV FPGA using Quartus II 12.0 and verified using SignalTAP II and Matlab. The simulation results show that the cryptoprocessors present a very good performance to carry out the scalar multiplication kP. In this case, the computation times of the multiplication kP using Lopez-Dahab, 2-NAF halve-and-add and 16-tNAF for Koblitz curves were 13.37 µs, 16.90 µs and 5.05 µs, respectively.

  4. Calculation of particulate dispersion in a design-basis tornadic storm from Westinghouse PFDL, Cheswick, Pennsylvania

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1978-07-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Westinghouse Plutonium Fuel Development Laboratory (PFDL) at Cheswick, Pennsylvania. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20-45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

  5. Evaluating comfort with varying temperatures: a graphic design tool

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.M. [Research Centre Habitat and Energy, Faculty of Architecture, Design and Urbanism, University of Buenos Aires, Ciudad Universitaria (Argentina)

    2002-07-01

    This paper considers the need to define comfort of indoor and outdoor spaces in relation to the daily variations of temperature. A graphical tool is presented, which indicates the daily swings of temperature, shown as a single point on a graph representing the average temperature and the maximum temperature swing. This point can be compared with the comfort zones for different activity levels, such as sedentary activity, sleeping, indoor and outdoor circulation according to the design proposals for different spaces. The graph allows the representation of climatic variables, the definition of comfort zones, the selection of bio climatic design resources and the evaluation of indoor temperatures, measured in actual buildings or obtained from computer simulations. The development of the graph is explained and examples given with special emphasis on the use of thermal mass. (author)

  6. Design and Development of a PC- Based temperature monitoring ...

    African Journals Online (AJOL)

    The design of the work involves a circuit that measures the surrounding temperature using appropriate sensors and the sensor output is then converted to digital signals after due processing and conditioning of the signals. There is also an interface circuit configured to make it compatible with the PC hardware. This design ...

  7. Design of PID temperature control system based on STM32

    Science.gov (United States)

    Zhang, Jianxin; Li, Hailin; Ma, Kai; Xue, Liang; Han, Bianhua; Dong, Yuemeng; Tan, Yue; Gu, Chengru

    2018-03-01

    A rapid and high-accuracy temperature control system was designed using proportional-integral-derivative (PID) control algorithm with STM32 as micro-controller unit (MCU). The temperature control system can be applied in the fields which have high requirements on the response speed and accuracy of temperature control. The temperature acquisition circuit in system adopted Pt1000 resistance thermometer as temperature sensor. Through this acquisition circuit, the monitoring actual temperature signal could be converted into voltage signal and transmitted into MCU. A TLP521-1 photoelectric coupler was matched with BD237 power transistor to drive the thermoelectric cooler (TEC) in FTA951 module. The effective electric power of TEC was controlled by the pulse width modulation (PWM) signals which generated by MCU. The PWM signal parameters could be adjusted timely by PID algorithm according to the difference between monitoring actual temperature and set temperature. The upper computer was used to input the set temperature and monitor the system running state via serial port. The application experiment results show that the temperature control system is featured by simple structure, rapid response speed, good stability and high temperature control accuracy with the error less than ±0.5°C.

  8. Extreme load alleviation using industrial implementation of active trailing edge flaps in a full design load basis

    DEFF Research Database (Denmark)

    Barlas, Athanasios; Pettas, Vasilis; Gertz, Drew Patrick

    2016-01-01

    The application of active trailing edge flaps in an industrial oriented implementation is evaluated in terms of capability of alleviating design extreme loads. A flap system with basic control functionality is implemented and tested in a realistic full Design Load Basis (DLB) for the DTU 10MW...

  9. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  10. Design and Implementation of High Precision Temperature Measurement Unit

    Science.gov (United States)

    Zeng, Xianzhen; Yu, Weiyu; Zhang, Zhijian; Liu, Hancheng

    2018-03-01

    Large-scale neutrino detector requires calibration of photomultiplier tubes (PMT) and electronic system in the detector, performed by plotting the calibration source with a group of designated coordinates in the acrylic sphere. Where the calibration source positioning is based on the principle of ultrasonic ranging, the transmission speed of ultrasonic in liquid scintillator of acrylic sphere is related to temperature. This paper presents a temperature measurement unit based on STM32L031 and single-line bus digital temperature sensor TSic506. The measurement data of the temperature measurement unit can help the ultrasonic ranging to be more accurate. The test results show that the temperature measurement error is within ±0.1°C, which satisfies the requirement of calibration source positioning. Take energy-saving measures, with 3.7V/50mAH lithium battery-powered, the temperature measurement unit can work continuously more than 24 hours.

  11. Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

    International Nuclear Information System (INIS)

    Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

    2013-01-01

    The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

  12. High temperature structure design for FBRs and analysis technology

    International Nuclear Information System (INIS)

    Iwata, Koji

    1986-01-01

    In the case of FBRs, the operation temperature exceeds 500 deg C, therefore, the design taking the inelastic characteristics of structural materials, such as plasticity and creep, into account is required, and the high grade and detailed evaluation of design is demanded. This new high temperature structure design technology has been advanced in respective countries taking up experimental, prototype and demonstration reactors as the targets. The development of FBRs in Japan was begun with the experimental reactor 'Joyo' which has been operated since 1977, and now, the prototype FBR 'Monju' of 280 MWe is under construction, which is expected to attain the criticality in 1992. In order to realize FBRs which can compete with LWRs through the construction of a demonstration FBR, the construction of large scale plants and the heightening of the economy and reliability are necessary. The features and the role of FBR structural design, the method of high temperature structure design and the trend of its standardization, the trend of the structural analysis technology for FBRs such as inelastic analysis, buckling analysis and fluid and structure coupled vibration analysis, the present status of structural analysis programs, and the subjects for the future of high temperature structure design are explained. (Kako, I.)

  13. Structural analysis technology for high-temperature design

    International Nuclear Information System (INIS)

    Greenstreet, W.L.

    1977-01-01

    Results from an ongoing program devoted to the development of verified high-temperature structural design technology applicable to nuclear reactor systems are described. The major aspects addressed by the program are (1) deformation behavior; (2) failure associated with creep rupture, brittle fracture, fatigue, creep-fatigue interactions, and crack propagation; and (3) the establishment of appropriate design criteria. This paper discusses information developed in the deformation behavior category. The material considered is type 304 stainless steel, and the temperatures range to 1100 0 F (593 0 C). In essence, the paper considers the ingredients necessary for predicting relatively high-temperature inelastic deformation behavior of engineering structures under time-varying temperature and load conditions and gives some examples. These examples illustrate the utility and acceptability of the computational methods identified and developed for prediting essential features of complex inelastic behaviors. Conditions and responses that can be encountered under nuclear reactor service conditions and invoked in the examples. (Auth.)

  14. Designing Robustness to Temperature in a Feedforward Loop Circuit

    OpenAIRE

    Sen, Shaunak; Kim, Jongmin; Murray, Richard M.

    2013-01-01

    Incoherent feedforward loops represent important biomolecular circuit elements capable of a rich set of dynamic behavior including adaptation and pulsed responses. Temperature can modulate some of these properties through its effect on the underlying reaction rate parameters. It is generally unclear how to design such a circuit where the properties are robust to variations in temperature. Here, we address this issue using a combination of tools from control and dynamical systems theory as wel...

  15. Design and Implementation of Temperature Controller for a Vacuum Distiller

    OpenAIRE

    Muslim, M. Aziz; N., Goegoes Dwi; F., Ahmad Salmi; R., Akhbar Prachaessardhi

    2014-01-01

    This paper proposed design and implementation of temperature controller for a vacuum distiller. The distiller is aimed to provide distillation process of bioethanol in nearly vacuum condition. Due to varying vacuum pressure, temperature have to be controlled by manipulating AC voltage to heating elements. Two arduino based control strategies have been implemented, PID control and Fuzzy Logic control. Control command from the controller was translated to AC drive using TRIAC based dimmer circu...

  16. Heat experiment design to estimate temperature dependent thermal properties

    International Nuclear Information System (INIS)

    Romanovski, M

    2008-01-01

    Experimental conditions are studied to optimize transient experiments for estimating temperature dependent thermal conductivity and volumetric heat capacity. A mathematical model of a specimen is the one-dimensional heat equation with boundary conditions of the second kind. Thermal properties are assumed to vary nonlinearly with temperature. Experimental conditions refer to the thermal loading scheme, sampling times and sensor location. A numerical model of experimental configurations is studied to elicit the optimal conditions. The numerical solution of the design problem is formulated on a regularization scheme with a stabilizer minimization without a regularization parameter. An explicit design criterion is used to reveal the optimal sensor location, heating duration and flux magnitude. Results obtained indicate that even the strongly nonlinear experimental design problem admits the aggregation of its solution and has a strictly defined optimal measurement scheme. Additional region of temperature measurements with allowable identification error is revealed.

  17. Design of temperature detection device for drum of belt conveyor

    Science.gov (United States)

    Zhang, Li; He, Rongjun

    2018-03-01

    For difficult wiring and big measuring error existed in the traditional temperature detection method for drum of belt conveyor, a temperature detection device for drum of belt conveyor based on Radio Frequency(RF) communication is designed. In the device, detection terminal can collect temperature data through tire pressure sensor chip SP370 which integrates temperature detection and RF emission. The receiving terminal which is composed of RF receiver chip and microcontroller receives the temperature data and sends it to Controller Area Network(CAN) bus. The test results show that the device meets requirements of field application with measuring error ±3.73 ° and single button battery can provide continuous current for the detection terminal over 1.5 years.

  18. Analysis of optimal design of low temperature economizer

    Science.gov (United States)

    Song, J. H.; Wang, S.

    2017-11-01

    This paper has studied the Off-design characteristic of low temperature economizer system based on thermodynamics analysis. Based on the data from one 1000 MW coal-fired unit, two modes of operation are contrasted and analyzed. One is to fix exhaust gas temperature and the other one is to take into account both of the average temperature difference and the exhaust gas temperature. Meanwhile, the cause of energy saving effect change is explored. Result shows that: in mode 1, the amount of decrease in coal consumption reduces from 1.11 g/kWh (under full load) to 0.54 g/kWh (under half load), and in mode 2, when the load decreases from 90% to 50%, the decrease in coal consumption reduces from 1.29 g/kWh to 0.84 g/kWh. From the result, under high load, the energy saving effect is superior, and under lower work load, energy saving effect declines rapidly when load is reduced. When load changes, the temperature difference of heat transfer, gas flow, the flue gas heat rejection and the waste heat recovery change. The energy saving effect corresponding changes result in that the energy saving effect under high load is superior and more stable. However, rational adjustment to the temperature of outlet gas can alleviate the decline of the energy saving effect under low load. The result provides theoretical analysis data for the optimal design and operation of low temperature economizer system of power plant.

  19. Criteria for calculating the efficiency of deep-pleated HEPA filters with aluminum separators during and after design basis accidents

    International Nuclear Information System (INIS)

    Bergman, W.; First, M.W.; Anderson, W.L.

    1995-01-01

    We have reviewed the literature on the performance of HEPA filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). This study is only applicable to the standard deep-pleated HEPA filter with aluminum separators as specified in ASME N509[1]. Other HEPA filter designs such as the mini-pleat and separatorless filters are not included in this study. The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be damaged and have a residual efficiency of 0%. There are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen. The estimation of the efficiency of the HEPA filters under DBA conditions involves three steps: (1) The filter pressure drop and environmental parameters are determined during and after the DBA, (2) Comparing the filter pressure drop to a set of threshold values above which the filter is damaged. There is a different threshold value for each combination of environmental parameters, and (3) Determining the filter efficiency. If the filter pressure drop is greater than the threshold value, the filter is damaged and is assigned 0% efficiency. If the pressure drop is less, then the filter is not damaged and the efficiency is determined from literature values of the efficiency at the environmental conditions

  20. Criteria for calculating the efficiency of deep-pleated HEPA filters with aluminum separators during and after design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    Bergman, W.; First, M.W.; Anderson, W.L. [Lawrence Livermore National Laboratory, CA (United States)] [and others

    1995-02-01

    We have reviewed the literature on the performance of HEPA filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). This study is only applicable to the standard deep-pleated HEPA filter with aluminum separators as specified in ASME N509[1]. Other HEPA filter designs such as the mini-pleat and separatorless filters are not included in this study. The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be damaged and have a residual efficiency of 0%. There are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen. The estimation of the efficiency of the HEPA filters under DBA conditions involves three steps: (1) The filter pressure drop and environmental parameters are determined during and after the DBA, (2) Comparing the filter pressure drop to a set of threshold values above which the filter is damaged. There is a different threshold value for each combination of environmental parameters, and (3) Determining the filter efficiency. If the filter pressure drop is greater than the threshold value, the filter is damaged and is assigned 0% efficiency. If the pressure drop is less, then the filter is not damaged and the efficiency is determined from literature values of the efficiency at the environmental conditions.

  1. Reactor similarity for plasma–material interactions in scaled-down tokamaks as the basis for the Vulcan conceptual design

    International Nuclear Information System (INIS)

    Whyte, D.G.; Olynyk, G.M.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Hartwig, Z.S.; Mumgaard, R.T.; Podpaly, Y.A.

    2012-01-01

    Highlights: ► Discussion of similarity scalings for reduced-size tokamaks. ► Proposal of a new set of scaling laws for divertor similarity. ► Discussion of how the new scaling provides fidelity to a reactor. ► The new scaling is used as the basis for the Vulcan conceptual design. - Abstract: Dimensionless parameter scaling techniques are a powerful tool in the study of complex physical systems, especially in tokamak fusion experiments where the cost of full-size devices is high. It is proposed that dimensionless similarity be used to study in a small-scale device the coupled issues of the scrape-off layer (SOL) plasma, plasma–material interactions (PMI), and the plasma-facing material (PFM) response expected in a tokamak fusion reactor. Complete similarity is not possible in a reduced-size device. In addition, “hard” technological limits on the achievable magnetic field and peak heat flux, as well as the necessity to produce non-inductive scenarios, must be taken into account. A practical approach is advocated, in which the most important dimensionless parameters are matched to a reactor in the reduced-size device, while relaxing those parameters which are far from a threshold in behavior. “Hard” technological limits are avoided, so that the reduced-size device is technologically feasible. A criticism on these grounds is offered of the “P/R” model, in which the ratio of power crossing the last closed flux surface (LCFS), P, to the device major radius, R, is held constant. A new set of scaling rules, referred to as the “P/S” scaling (where S is the LCFS area) or the “PMI” scaling, is proposed: (i) non-inductive, steady-state operation; (ii) P is scaled with R 2 so that LCFS areal power flux P/S is constant; (iii) magnetic field B constant; (iv) geometry (elongation, safety factor q * , etc.) constant; (v) volume-averaged core density scaled as n≈n ¯ e ∼R −2/7 ; and (vi) ambient wall material temperature T W,0 constant. It is

  2. Inferring seawater temperature over the past 2,500 years in the Southern California Bight on the basis of brachiopods

    Science.gov (United States)

    Tomašových, Adam; Müller, Tamás; Kidwell, Susan M.

    2017-04-01

    Use of calcite δ18O in brachiopod shells in assessing past variations in seawater temperature remains poorly constrained in the absence of other methods due to vital effects and unknown variations in seawater density, salinity. Here, in order to evaluate past changes in seawater temperature of mainland shelf habitats off the Southern California Bight over the past 2,500 years, we analyze δ18O and Mg/Ca ratio of dead shells of the terebratulid brachiopod Laqueus erythraeus collected at 60-80 m water depths and age-dated by radiocarbon-calibrated amino acid racemization. These dead Holocene shells show excellent preservation (Mn concentrations < 10 ppm and Sr concentrations above 800 ppm). Although historical changes in sea-surface temperature in the southern California Bight were inferred on the basis of alkenones and δ18O in of planktonic foraminifers, temperature history of deeper shelf below storm wave base in this region remains unclear. First, we investigate thermal sensitivity of Mg/Ca ratio (using Laser Ablation Inductively Coupled Plasma Mass Spectrometry and wavelength-dispersive spectrometry) in the terebratulid brachiopod Laqueus erythraeus (collected in 1994 at Santa Catalina Island at 116 m water depth). At this depth, annual temperature range is relatively small (between 9-11°C), although at times of El Nino events in 1982-1983, 1986-1987, and 1992-1993, monthly temperature attained 13 °C. We find that δ18O measured along a growth profile of a shell precipitated in oxygen isotopic equilibrium with ambient seawater, and maxima in Mg/Ca ratio coincide with minima in δ18O, suggesting that fluctuations in Mg/Ca ratio trace temperature fluctuations, as observed also in other brachiopod species. Second, preliminary observations of Holocene shells show that Mg/Ca ratios show centennial-scale fluctuations but on average remain remarkably constant, with minima and maxima staying within intra-shell seasonal variations captured by extant specimens

  3. Alternative designs of high-temperature superconducting synchronous generators

    OpenAIRE

    Goddard, K. F.; Lukasik, B.; Sykulski, J. K.

    2010-01-01

    This paper discusses the different possible designs of both cored and coreless superconducting synchronous generators using high-temperature superconducting (HTS) tapes, with particular reference to demonstrators built at the University of Southampton using BiSCCO conductors. An overview of the electromagnetic, thermal, and mechanical issues is provided, the advantages and drawbacks of particular designs are highlighted, the need for compromises is explained, and practical solutions are offer...

  4. Structural Basis for Escape of Human Astrovirus from Antibody Neutralization: Broad Implications for Rational Vaccine Design

    Energy Technology Data Exchange (ETDEWEB)

    Bogdanoff, Walter A.; Perez, Edmundo I.; López, Tomás; Arias, Carlos F.; DuBois, Rebecca M. (UNAM-Mexico); (UCSC)

    2017-10-25

    ABSTRACT

    Human astroviruses are recognized as a leading cause of viral diarrhea worldwide in children, immunocompromised patients, and the elderly. There are currently no vaccines available to prevent astrovirus infection; however, antibodies developed by healthy individuals during previous infection correlate with protection from reinfection, suggesting that an effective vaccine could be developed. In this study, we investigated the molecular mechanism by which several strains of human astrovirus serotype 2 (HAstV-2) are resistant to the potent HAstV-2-neutralizing monoclonal antibody PL-2 (MAb PL-2). Sequencing of the HAstV-2 capsid genes reveals mutations in the PL-2 epitope within the capsid's spike domain. To understand the molecular basis for resistance from MAb PL-2 neutralization, we determined the 1.35-Å-resolution crystal structure of the capsid spike from one of these HAstV-2 strains. Our structure reveals a dramatic conformational change in a loop within the PL-2 epitope due to a serine-to-proline mutation, locking the loop in a conformation that sterically blocks binding and neutralization by MAb PL-2. We show that mutation to serine permits loop flexibility and recovers MAb PL-2 binding. Importantly, we find that HAstV-2 capsid spike containing a serine in this loop is immunogenic and elicits antibodies that neutralize all HAstV-2 strains. Taken together, our results have broad implications for rational selection of vaccine strains that do not contain prolines in antigenic loops, so as to elicit antibodies against diverse loop conformations.

    IMPORTANCEHuman astroviruses (HAstVs) infect nearly every person in the world during childhood and cause diarrhea, vomiting, and fever. In this study, we investigated how several strains of HAstV are resistant to a virus-neutralizing monoclonal antibody. We determined the crystal structure of the capsid protein spike domain from one of these HAstV strains and found that

  5. Application of Combined Sustained and Cyclic Loading Test Results to Alloy 617 Elevated Temperature Design Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yanli [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jetter, Robert I [Global Egineering and Technology, LLC, Coral Gables, FL (United States); Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-08-25

    Alloy 617 is a reference structural material for very high temperature components of advanced-gas cooled reactors with outlet temperatures in the range of 900-950°C . In order for designers to be able to use Alloy 617 for these high temperature components, Alloy 617 has to be approved for use in Section III (the nuclear section) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. A plan has been developed to submit a draft code for Alloy 617 to ASME Section III by 2015. However, the current rules in Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200°F (650°C). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep deformation, which is the basis for the current simplified rules. This temperature, 1200 °F, is well below the temperature range of interest for this material in High Temperature Gas Cooled Reactor (HTGR) applications. The only current alternative is, thus, a full inelastic analysis which requires sophisticated material models which have been formulated but not yet verified. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods and which are expected to be applicable to very high temperatures.

  6. Implementation of an Industrial-Based Case Study as the Basis for a Design Project in an Introduction to Mechanical Design Course

    Science.gov (United States)

    Lackey, Ellen

    2011-01-01

    The purpose of this paper is to discuss the implementation of an industrial-based case study as the basis for a design project for the Spring 2009 Introduction to Mechanical Design Course at the University of Mississippi. Course surveys documented the lack of student exposure in classes to the types of projects typically experienced by engineers…

  7. Junction Temperature Aware Energy Efficient Router Design on FPGA

    DEFF Research Database (Denmark)

    Thind, Vandana; Sharma, Shivani; Minwer, M H

    2015-01-01

    Energy, Power and efficiency are very much related to each other. To make any system efficient, Power consumed by it must be minimized or we can say that power dissipation should be less. In our research we tried to make a energy efficient router design on FPGA by varying junction temperature...

  8. Lower-Temperature Invert Design For Diffusion Barrier

    International Nuclear Information System (INIS)

    Bruce Stanley

    2001-01-01

    The objective of this analysis is to advance the state of the subsurface facilities design to primarily support the ''Yucca Mountain Science and Engineering Report'' (DOE 2001) and to also support the preparation and revision of System Description Document's Section 2 system descriptions (CRWMS M and O 2001, pp. 9 and 11). The results may also eventually support the License Application (CRWMS M and O 2001, p. 3). The Performance Assessment Department will be the primary user of the information generated and will be used in abstraction modeling for the lower-temperature scenario (CRWMS M and O 200 1, p. 27). This analysis will evaluate the invert relative to the lower- and higher-temperature conditions in accordance with the primary tasks below. Invert design is a major factor in allowing water entering the drift to pass freely and enter the drift floor without surface ponding and in limiting diffusive transport into the host rock. Specific cost effective designs will be conceptualized under the new lower-temperature conditions in this analysis. Interfacing activities and all aspects of Integrated Safety Management and Nuclear Culture principles are included in this work scope by adhering to the respective principles during this design activity and by incorporating safety into the design analysis (CRWMS M and O 2001, p. 8). Primary tasks of this analysis include identifying available design information from existing sources on the invert as a diffusive barrier, developing concepts that reduce the amount steel, and developing other design features that accommodate both lower- and higher-temperature operating modes (CRWMS M and O 2001, p.16)

  9. Extreme load alleviation using industrial implementation of active trailing edge flaps in a full design load basis

    OpenAIRE

    Barlas, Athanasios; Pettas, Vasilis; Gertz, Drew Patrick; Aagaard Madsen , Helge

    2016-01-01

    The application of active trailing edge flaps in an industrial oriented implementation is evaluated in terms of capability of alleviating design extreme loads. A flap system with basic control functionality is implemented and tested in a realistic full Design Load Basis (DLB) for the DTU 10MW Reference Wind Turbine (RWT) model and for an upscaled rotor version in DTU's aeroelastic code HAWC2. The flap system implementation shows considerable potential in reducing extreme loads in components o...

  10. Cellular basis of morphological variation and temperature-related plasticity in Drosophila melanogaster strains with divergent wing shapes.

    Science.gov (United States)

    Torquato, Libéria Souza; Mattos, Daniel; Matta, Bruna Palma; Bitner-Mathé, Blanche Christine

    2014-12-01

    Organ shape evolves through cross-generational changes in developmental patterns at cellular and/or tissue levels that ultimately alter tissue dimensions and final adult proportions. Here, we investigated the cellular basis of an artificially selected divergence in the outline shape of Drosophila melanogaster wings, by comparing flies with elongated or rounded wing shapes but with remarkably similar wing sizes. We also tested whether cellular plasticity in response to developmental temperature was altered by such selection. Results show that variation in cellular traits is associated with wing shape differences, and that cell number may play an important role in wing shape response to selection. Regarding the effects of developmental temperature, a size-related plastic response was observed, in that flies reared at 16 °C developed larger wings with larger and more numerous cells across all intervein regions relative to flies reared at 25 °C. Nevertheless, no conclusive indication of altered phenotypic plasticity was found between selection strains for any wing or cellular trait. We also described how cell area is distributed across different intervein regions. It follows that cell area tends to decrease along the anterior wing compartment and increase along the posterior one. Remarkably, such pattern was observed not only in the selected strains but also in the natural baseline population, suggesting that it might be canalized during development and was not altered by the intense program of artificial selection for divergent wing shapes.

  11. Evaluation of the effectiveness of grouting curtain on the basis of the analysis of groundwater temperature fluctuations behind the dam

    Directory of Open Access Journals (Sweden)

    Orekhov Vyacheslav Valentinovich

    2015-04-01

    Full Text Available In the article the authors considered the technique of evaluating the performance of a grouting curtain basing on the analysis of mathematical forecasting and regular measurements of water temperature in the reservoir and in the rock mass behind the dam. The initial data for the solution of heat transfer problem are the rate of filtration, obtained from the solution of the stationary problem of filtration, and the experimental factor, generalizing thermophysical properties of rocks. For calculating the period of time from to the change of the water temperature in the reservoir till the change of water temperature at the reference point of the rock mass a computer program was designed, which allows defining the path and time of filtration from the reservoir to the reference point in the rock mass with the help of the reverse conversion on flow lines. The calculation was carried out from the point in question in the rock mass till the crossing paths of filtration with the bottom boundary of the reservoir. As an example, we present the results of computational studies of filtration and temperature regimes in the rock foundation of a concrete dam at the design work of the grouting curtain and in case of the presence of pervious area. The calculations were performed with a time step dt = 2 days. At each time step, with account of water motion along the lines of the current through the rock mass, the previous position of the reference points in space has been determined, for which the value of the velocity vector of filtration field was corrected. In the first case, the motion of water from the reservoir was carried out in the circumvention of the grouting curtain. In the second case, the motion of water took place from the reservoir through the permeable portion of the grouting curtain. The change of the water temperature during its seepage from the water reservoir through permeable area of grouting curtain because of conductive heat transmission in

  12. Embedded DAQ System Design for Temperature and Humidity Measurement

    Directory of Open Access Journals (Sweden)

    Tarique Rafique Memon

    2016-05-01

    Full Text Available In this work, we have proposed a cost effective DAQ (Data Acquisition system design useful for local industries by using user friendly LABVIEW (Laboratory Virtual Instrumentation Electronic Workbench. The proposed system can measure and control different industrial parameters which can be presented in graphical icon format. The system design is proposed for 8-channels, whereas tested and recorded for two parameters i.e. temperature and RH (Relative Humidity. Both parameters are set as per upper and lower limits and controlled using relays. Embedded system is developed using standard microcontroller to acquire and process the analog data and plug-in for further processing using serial interface with PC using LABVIEW. The designed system is capable of monitoring and recording the corresponding linkage between temperature and humidity in industrial unit's and indicates the abnormalities within the process and control those abnormalities through relays

  13. Embedded DAQ System Design for Temperature and Humidity Measurement

    International Nuclear Information System (INIS)

    Memon, T.R.

    2013-01-01

    In this work, we have proposed a cost effective DAQ (Data Acquisition) system design useful for local industries by using user friendly LABVIEW (Laboratory Virtual Instrumentation Electronic Workbench). The proposed system can measure and control different industrial parameters which can be presented in graphical icon format. The system design is proposed for 8-channels, whereas tested and recorded for two parameters i.e. temperature and RH (Relative Humidity). Both parameters are set as per upper and lower limits and controlled using relays. Embedded system is developed using standard microcontroller to acquire and process the analog data and plug-in for further processing using serial interface with PC using LABVIEW. The designed system is capable of monitoring and recording the corresponding linkage between temperature and humidity in industrial unit's and indicates the abnormalities within the process and control those abnormalities through relays. (author)

  14. Technical basis for the ITER final design report, cost review and safety analysis (FDR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

    The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement. Refs, figs, tabs

  15. Technical basis for the ITER final design report, cost review and safety analysis (FDR)

    International Nuclear Information System (INIS)

    1998-01-01

    The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement

  16. Applying dynamic mold temperature control to cosmetic package design

    Directory of Open Access Journals (Sweden)

    Hsiao Shih-Wen

    2017-01-01

    Full Text Available Owing to the fashion trend and the market needs, this study developed the eco-cushion compact. Through the product design and the advanced process technology, many issues have improved, for instance, the inconvenience of transportation, the lack of multiuse capability, the increase of costs, and the low yield rate. The eco-cushion compact developed in this study was high quality, low cost, and meets the requirements of the eco market. The study aimed at developing a reusable container. Dynamic mold temperature control was introduced in the injection modeling process. The innovation in the product was its multi-functional formula invention, eco-product design, one-piece powder case design, and multifunctional design in the big powder case, mold flow and development of dynamic mold temperature control. Finally, through 3D drawing and modeling, and computer assistance for mold flow and verification to develop and produce models. During the manufacturing process, in order to solve the problems of tightness and warping, development and manufacture of dynamic mold temperature control were introduced. This decreased the injection cycle and residual stress, and deformation of the products has reduced to less than 0.2 mm, and the air tightness increased. In addition, air leakage was less than 2% and the injection cycle decreased to at least 10%. The results of the study can be extended and applied on the future design on cosmetic package and an alternative can be proposed to solve the problems of air tightness and warping. In this study, dynamic mold temperature control is considered as a design with high price-performance ratio, which can be adopted on industrial application for practical benefit and improvement.

  17. RESEARCH OF HEAT-RESISTANT CONCRETE ON THE BASIS OF BASALT FILLER FOR CONCRETING OF METAL DESIGNS

    Directory of Open Access Journals (Sweden)

    R. M. Curbanov

    2013-01-01

    Full Text Available Expediency of use of heat-resistant concrete locates in article on the basis of a basalt filler. It is thin a ground additive promotes increase in power of internal friction between material particles. With increase in power of internal friction between particles viscosity knitting increases and as a result ryazmyagcheniye temperature under loading increases and fire resistance of a material increases

  18. Design and modeling of low temperature solar thermal power station

    International Nuclear Information System (INIS)

    Shankar Ganesh, N.; Srinivas, T.

    2012-01-01

    Highlights: ► The optimum conditions are different for efficiency and power conditions. ► The current model works up to a maximum separator temperature of 150 °C. ► The turbine concentration influences the high pressure. ► High solar beam radiation and optimized cycle conditions give low collector cost. -- Abstract: During the heat recovery in a Kalina cycle, a binary aqua–ammonia mixture changes its state from liquid to vapor, the more volatile ammonia vaporizes first and then the water starts vaporization to match temperature profile of the hot fluid. In the present work, a low temperature Kalina cycle has been investigated to optimize the heat recovery from solar thermal collectors. Hot fluid coming from solar parabolic trough collector with vacuum tubes is used to generate ammonia rich vapor in a boiler for power generation. The turbine inlet conditions are optimized to match the variable hot fluid temperature with the intermittent nature of the solar radiation. The key parameters discussed in this study are strong solution concentration, separator temperature which affects the hot fluid inlet temperature and turbine ammonia concentration. Solar parabolic collector system with vacuum tubes has been designed at the optimized power plant conditions. This work can be used in the selection of boiler, separator and turbine conditions to maximize the power output as well as efficiency of power generation system. The current model results a maximum limit temperature for separator as 150 °C at the Indian climatic conditions. A maximum specific power of 105 kW per kg/s of working fluid can be obtained at 80% of strong solution concentration with 140 °C separator temperature. The corresponding plant and cycle efficiencies are 5.25% and 13% respectively. But the maximum efficiencies of 6% and 15% can be obtained respectively for plant and Kalina cycle at 150 °C of separator temperature.

  19. Design for a low temperature ion implantation and luminescence cryostat

    International Nuclear Information System (INIS)

    Noonan, J.R.; Kirkpatrick, C.G.; Myers, D.R.; Streetman, B.G.

    1976-01-01

    Several simple design changes of a conventional liquid helium optical Dewar can significantly improve the cryostat's versatility for use in low temperature particle irradiation. A bellows assembly provides precise sample positioning and allows convenient access for electrical connections. A heat exchanger consisting of thin walled tubing with a 'goose neck' bend provides a simple, effective means of cooling the sample as well as excellent thermal isolation of the sample holder from the coolant reservoir during controlled anneals. The addition of a vane-type vacuum valve, optical windows, and a rotatable tailpiece facilitates the study of optical properties of materials following low temperature ion implantation. (author)

  20. Mathematical modelling of steam generator and design of temperature regulator

    Energy Technology Data Exchange (ETDEWEB)

    Bogdanovic, S.S. [EE Institute Nikola Tesla, Belgrade (Yugoslavia)

    1999-07-01

    The paper considers mathematical modelling of once-through power station boiler and numerical algorithm for simulation of the model. Fast and numerically stable algorithm based on the linearisation of model equations and on the simultaneous solving of differential and algebraic equations is proposed. The paper also presents the design of steam temperature regulator by using the method of projective controls. Dynamic behaviour of the system closed with optimal linear quadratic regulator is taken as the reference system. The desired proprieties of the reference system are retained and solutions for superheated steam temperature regulator are determined. (author)

  1. Technical basis for the ITER detailed design report, cost review and safety analysis (DDR)

    International Nuclear Information System (INIS)

    1997-01-01

    The ITER Detailed Design Report (DDR), Cost Review and Safety Analysis is the 3rd major milestone representing the progress made in the ITER Engineering Design Activities. With the approval of the Interim Design Report (IDR), it has been possible to freeze the main concepts and system approaches for ITER and to develop the design in more detail for the individual components and sub-systems. This report, although designed to be fully understandable as a separate document, focusses particularly on the main changes since the IDR

  2. Technical basis for the ITER detailed design report, cost review and safety analysis (DDR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The ITER Detailed Design Report (DDR), Cost Review and Safety Analysis is the 3rd major milestone representing the progress made in the ITER Engineering Design Activities. With the approval of the Interim Design Report (IDR), it has been possible to freeze the main concepts and system approaches for ITER and to develop the design in more detail for the individual components and sub-systems. This report, although designed to be fully understandable as a separate document, focusses particularly on the main changes since the IDR. Refs, figs, tabs

  3. Design basis for the copper/steel canister. Stage five. Final report

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    1999-05-01

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste in the Swedish Program, has been studied by the present author from the points of view of choice of materials, manufacturing technology and quality assurance. This report describes the observations on progress that has been made between May-1-1998 and April-30-1999 and the result of further literature studies. Cast steel has been rejected in favour of cast iron as a candidate material for the load bearing liner. The nodular iron that was selected has been the subject of casting trials at several foundries. Early trials, using uphill feeding, met with limited success owing to difficulties feeding during solidification. Lessons from this trial led to a modification to the casting design to include extra cores that have the effect of reducing the need for feeding in the heaviest sections. Results using the new design and direct (downhill) casting are very promising. Castings appear to be sound and mechanical test results cast-on bars are within specification. Tensile test results from specimens cut from the casting have reduced ductility compared with the cast-on bars and this may be evidence of microstructural variations within the casting. The material specified for the overpack is OF (Oxygen Free) copper with 50 ppm of phosphorus added. Concentration limits have now been placed on impurity elements which are below those allowed in the OF specification. All current trials are using material from Outokompu produced from cathode on their OF(E) line, which delivers total impurity levels of less than 30 ppm excluding silver and phosphorus. The phosphorus addition is made using a master alloy added to the launder and this does not give good control of phosphorus level either within or between castings. Phosphorus is added to improve creep rates and creep strain to failure. The level is limited to 50 ppm in order to avoid difficulties, which it might

  4. Design basis for the copper/steel canister. Stage five. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, W.H. [Meadow End Farm, Farnham (United Kingdom)

    1999-05-01

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste in the Swedish Program, has been studied by the present author from the points of view of choice of materials, manufacturing technology and quality assurance. This report describes the observations on progress that has been made between May-1-1998 and April-30-1999 and the result of further literature studies. Cast steel has been rejected in favour of cast iron as a candidate material for the load bearing liner. The nodular ironthat was selected has been the subject of casting trials at several foundries. Early trials, using uphill feeding, met with limited success owing to difficulties feeding during solidification. Lessons from this trial led to a modification to the casting design to include extra cores that have the effect of reducing the need for feeding in the heaviest sections. Results using the new design and direct (downhill) casting are very promising. Castings appear to be sound and mechanical test results cast-on bars are within specification. Tensile test results from specimens cut from the casting have reduced ductility compared with the cast-on bars and this may be evidence of microstructural variations within the casting. The material specified for the overpack is OF (Oxygen Free) copper with 50 ppm of phosphorus added. Concentration limits have now been placed on impurity elements which are below those allowed in the OF specification. All current trials are using material from Outokompu produced from cathode on their OF(E) line, which delivers total impurity levels of less than 30 ppm excluding silver and phosphorus. The phosphorus addition is made using a master alloy added to the launder and this does not give good control of phosphorus level either within or between castings. Phosphorus is added to improve creep rates and creep strain to failure. The level is limited to 50 ppm in order to avoid difficulties, which it might

  5. Development of Very High Temperature Reactor Design Technology

    International Nuclear Information System (INIS)

    Lee, Won Jae; Noh, J. M.; Kim, K. S.

    2007-05-01

    To develop design technologies for the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature for an efficient hydrogen production, key studies were performed, which include evaluation technology for the performance and safety of VHTR, and development of design and analysis codes. First, to evaluate the performance of VHTR, a series of analyses has been carried out for core characteristics at 950 .deg. C, cooled-vessel adopting internal flow path through the graphite structure, compact heat exchanger with periodic channel configuration, intermediate loop system, risk/performance-informed method, and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC, safety evaluation of both VHTR and RCCS has been also performed. In addition, prototype codes have been developed for a nuclear design, system loop design, system performance analysis, air-ingress accident analysis, fission product/tritium transport analysis, graphite structure seismic analysis and hydrogen explosion analysis, and they are being verified and validated through a lot of international collaborations

  6. A model for personal life project design on the basis of vocational guidance

    Directory of Open Access Journals (Sweden)

    Isaac Geovanni Mendoza Cedeño

    2015-09-01

    Full Text Available This article is intended to propose a model for developing vocational education personal life projects and its corresponding theoretical foundations. Therefore, the argument is undertaken on the model developed from philosophical views, epistemological, pedagogical, sociological, and psychological axiological, and a model is provided as a basis for the development of vocational education strategy for personal life project development contributing effectively to the development of responsible autonomy and high school students.

  7. Retrofitting a spent fuel pool spray system for alternative cooling as a strategy for beyond design basis events

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph; Vujic, Zoran [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2017-06-15

    Due to requirements for nuclear power plants to withstand beyond design basis accidents, including events such as happened in 2011 in the Fukushima Daiichi Nuclear Power Plant in Japan, alternative cooling of spent fuel is needed. Alternative spent fuel cooling can be provided by a retrofitted spent fuel pool spray system based on the AP1000 plant design. As part of Krsko Nuclear Power Plant's Safety Upgrade Program, Krsko Nuclear Power Plant decided on, and Westinghouse successfully designed a retrofit of the AP1000 {sup registered} plant spent fuel pool spray system to provide alternative spent fuel cooling.

  8. Approach to developing a ground-motion design basis for facilities important to safety at Yucca Mountain

    International Nuclear Information System (INIS)

    King, J.L.

    1990-01-01

    This paper discusses a methodology for developing a ground-motion design basis for prospective facilities at Yucca Mountain that are important to safety. The methodology utilizes a guasi-deterministic construct called the 10,000-year cumulative-slip earthquake that is designed to provide a conservative, robust, and reproducible estimate of ground motion that has a one-in-ten chance of occurring during the preclosure period. This estimate is intended to define a ground-motion level for which the seismic design would ensure minimal disruption to operations engineering analyses to ensure safe performance are included

  9. Development of Pupils Picture Aesthetic Competences on the Basis of IT-didactic Designs of Digital Picture Production

    DEFF Research Database (Denmark)

    Rasmussen, Helle

    : The research method refers to Design Based Research, since the project is based on a design theoretical view of learning. (Cobb et. All 2003, Van den Akker 2006, Collins 2004). Learning is here to be understood as “a sign producing activity in a specific situation within an institutional framing”, which makes...... Education” (English Title), The Danish University of Education Cobb, P. et al. (2003): “Design Experiments in Educational Research” in “Educational Researcher”, vol. 32, no. 1. Collins, Allan et. al. (2004): “Design Research: Theoretical and Metodological Issuses” in “Journal of the Learning Sciences”, Vol...... Competences on the Basis of IT-didactic Designs of Digital Picture Production Proposal information: The topic for this presentation is an ongoing investigation of the connection between the learning outcome of digital picture production and IT-didactic designs, and it refers to a Ph.D.-project in progress...

  10. A design basis for the development of advanced CANDU control centres

    International Nuclear Information System (INIS)

    Feher, M.P.; Davey, E.C.; Lupton, L.R.

    1995-01-01

    The basic design for current CANDU control centres was established in the early 1970's. Plants constructed since then have, for the most part, retained the same basic design. Several factors have led to the need to re-examine CANDU control centre design for plants to be built beyond the year 2000. These factors include the changing roles and responsibilities for the operations staff, an improved understanding of operational issues associated with supervisory control, an improved understanding of human error in operational situations, the opportunity for improved plant performance through the introduction of new technologies, and marketing pressures. This paper describes the proposed design bases for the development of advanced control centres to be implemented in CANDU plants beyond the year 2000. Four areas have been defined covering design goals, design principles, operational bases, and plant functional bases. (author)

  11. A design basis for the development of advanced CANDU control centres

    Energy Technology Data Exchange (ETDEWEB)

    Feher, M P; Davey, E C; Lupton, L R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The basic design for current CANDU control centres was established in the early 1970`s. Plants constructed since then have, for the most part, retained the same basic design. Several factors have led to the need to re-examine CANDU control centre design for plants to be built beyond the year 2000. These factors include the changing roles and responsibilities for the operations staff, an improved understanding of operational issues associated with supervisory control, an improved understanding of human error in operational situations, the opportunity for improved plant performance through the introduction of new technologies, and marketing pressures. This paper describes the proposed design bases for the development of advanced control centres to be implemented in CANDU plants beyond the year 2000. Four areas have been defined covering design goals, design principles, operational bases, and plant functional bases. (author).

  12. Air temperature determination inside residual heat removal pump room of Angra-1 nuclear power plant after a design basic accident

    International Nuclear Information System (INIS)

    Siniscalchi, Marcio Rezende

    2005-01-01

    This work develops heat transfer theoretical models for determination of air temperature inside the Residual Heat Removal Pump Room of Angra 1 Nuclear Power Plant after a Design Basis Accident without forced ventilation. Two models had been developed. The differential equations are solved by analytical methods. A software in FORTRAN language are developed for simulations of temperature inside rooms for different geometries and materials. (author)

  13. Design manual. [High temperature heat pump for heat recovery system

    Energy Technology Data Exchange (ETDEWEB)

    Burch, T.E.; Chancellor, P.D.; Dyer, D.F.; Maples, G.

    1980-01-01

    The design and performance of a waste heat recovery system which utilizes a high temperature heat pump and which is intended for use in those industries incorporating indirect drying processes are described. It is estimated that use of this heat recovery system in the paper, pulp, and textile industries in the US could save 3.9 x 10/sup 14/ Btu/yr. Information is included on over all and component design for the heat pump system, comparison of prime movers for powering the compressor, control equipment, and system economics. (LCL)

  14. Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio

    2002-10-01

    An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

  15. Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2002-10-01

    An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

  16. Shielding design study of the demonstration fast breeder reactor. 2. Shielding design on the basis of the JASPER analysis

    International Nuclear Information System (INIS)

    Suzuoki, Zenro; Tabayashi, Masao; Handa, Hiroyuki; Iida, Masaaki; Takemura, Morio

    2000-01-01

    Conceptual shielding design has been performed for the Demonstration Fast Breeder Reactor (DFBR) to achieve further optimization and reduction of the plant construction cost. The design took into account its implications in overall plant configuration such as reduction of shields in the core, adoption of fission gas plenum in the lower portion of fuel assemblies, and adoption of gas expansion modules. Shielding criteria applied for the design are to secure fast neutron fluence on in-vessel structures as well as responses of the nuclear instrumentation system and to restrict secondary sodium activation. The design utilized the cross sections and the one- and two-dimensional discrete ordinates transport codes, whose verification had been performed by the JASPER experiment analysis. Correction factors yielded by the JASPER analysis were applied to the design calculations to obtain design values with improved accuracy. Design margins, which are defined by the ratios of the design criteria to the design values, were more than two for all shielding issues of interest, showing the adequacy of the shielding design of the DFBR. (author)

  17. THEORETICAL CONTENT FOUNDATION OF PREPARING TEACHERS OF LABOR TRAINING TO TEACHING BASIS OF DESIGN

    Directory of Open Access Journals (Sweden)

    Ihor Savenko

    2016-06-01

    Full Text Available In the article the basic principles of shaping the content of training future teachers of technology education to realize design activity in schools have been considered. Formation of the basic core knowledge, abilities and skills of future teachers should be provided in the fundamental scientific, general and cultural and professional training that is subjected to certain principles of pedagogical design. Design education as a powerful educative potential aimed at designing and providing humanitarian and cultural orientation of a student has been revealed. The design is a valued feature of professional education of future teachers of labor studies and promotes the development of a special type of culture and thinking that directs to educational values and determines the individual educational strategy of professional development.

  18. Development of very high temperature reactor design technology

    International Nuclear Information System (INIS)

    Lee, Won Jae; Noh, Jan Man

    2012-04-01

    or an efficient production of nuclear hydrogen, the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature and the interfacing system for the hydrogen production are required. We have developed various evaluation technologies for the performance and safety of VHTR through the accomplishment of this project. First, to evaluate the performance of VHTR, a series of analyses has been performed such as core characteristics at 950 .deg. C, applicability of cooled-vessel, intermediate loop system and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC and the analysis of the risk/performance-informed method, VHTR safety evaluation has been also performed. In addition, various design analysis codes have been developed for a nuclear design, system loop design, system performance analysis, fission product/tritium transport analysis, core thermo-fluid analysis, system layout analysis, graphite structure seismic analysis and hydrogen exposion analysis, and they are being verified and validated through a lot of international collaborations

  19. A new electrodynamic balance design for low temperature studies

    Science.gov (United States)

    Tong, H.-J.; Ouyang, B.; Pope, F. D.; Kalberer, M.

    2014-07-01

    In this paper we describe a newly designed cold electrodynamic balance (CEDB) system, which was built to study the evaporation kinetics and freezing properties of supercooled water droplets. The temperature of the CEDB chamber at the location of the levitated water droplet can be controlled in the range: -40 to +40 °C, which is achieved using a combination of liquid nitrogen cooling and heating by positive temperature coefficient heaters. The measurement of liquid droplet radius is obtained by analyzing the Mie elastic light scattering from a 532 nm laser. The Mie scattering signal was also used to characterize and distinguish droplet freezing events; liquid droplets produce a regular fringe pattern whilst the pattern from frozen particles is irregular. The evaporation rate of singly levitated water droplets was calculated from time resolved measurements of the radii of evaporating droplets and a clear trend of the evaporation rate on temperature was measured. The statistical freezing probabilities of aqueous pollen extracts (pollen washing water) are obtained in the temperature range: -4.5 to -40 °C. It was found that that pollen washing water from water birch (Betula fontinalis occidentalis) pollen can act as ice nuclei in the immersion freezing mode at temperatures as warm as -22.45 (±0.65) °C.

  20. Development of a design basis tornado and structural design criteria for the Nevada Test Site, Nevada. Final report

    International Nuclear Information System (INIS)

    McDonald, J.R.; Minor, J.E.; Mehta, K.C.

    1975-06-01

    In order to evaluate the ability of critical facilities at the Nevada Test Site to withstand the possible damaging effects of extreme winds and tornadoes, parameters for the effects of tornadoes and extreme winds and structural design criteria for the design and evaluation of structures were developed. The meteorological investigations conducted are summarized, and techniques used for developing the combined tornado and extreme wind risk model are discussed. The guidelines for structural design include methods for calculating pressure distributions on walls and roofs of structures and methods for accommodating impact loads from wind-driven missiles. Calculations for determining the design loads for an example structure are included

  1. The Concept of Fashion Design on the Basis of Color Coordination Using White LED Lighting

    Science.gov (United States)

    Mizutani, Yumiko; Taguchi, Tsunemasa

    This thesis focuses on the development of fashion design, especially a dress coordinated with White LED Lighting (=LED). As for the design concept a fusion of the advanced science and local culture was aimed for. For such a reason this development is a very experimental one. Here in particular I handled an Imperial Court dinner dress for the last Japanese First Lady, Mrs. Akie Abe who wore it at the Imperial Court dinner for the Indonesian First Couple held on November 2006 to. This dress made by Prof. T. Taguchi and I open up a new field in the dress design.

  2. Probabilistic risk assessment for the Los Alamos Meson Physics Facility worst-case design-basis accident

    International Nuclear Information System (INIS)

    Sharirli, M.; Butner, J.M.; Rand, J.L.; Macek, R.J.; McKinney, S.J.; Roush, M.L.

    1992-01-01

    This paper presents results from a Los Alamos National Laboratory Engineering and Safety Analysis Group assessment of the worse-case design-basis accident associated with the Clinton P. Anderson Meson Physics Facility (LAMPF)/Weapons Neutron Research (WNR) Facility. The primary goal of the analysis was to quantify the accident sequences that result in personnel radiation exposure in the WNR Experimental Hall following the worst-case design-basis accident, a complete spill of the LAMPF accelerator 1L beam. This study also provides information regarding the roles of hardware systems and operators in these sequences, and insights regarding the areas where improvements can increase facility-operation safety. Results also include confidence ranges to incorporate combined effects of uncertainties in probability estimates and importance measures to determine how variations in individual events affect the frequencies in accident sequences

  3. The neuroscientific basis of successful design how emotions and perceptions matter

    CERN Document Server

    Maiocchi, Marco

    2015-01-01

    The term “design” today encompasses attributes of artifacts that go beyond their intended functions, imbuing them with new meanings. Those meanings are deeply related to the emotions perceived by the users. This book investigates the findings deriving from the neurosciences that are relevant to design. Drawing upon up-to-date neuroscientific knowledge, the authors define what an emotion is, examine the relationship between perceptions and emotions, and discuss the role of metaphoric communication. Particular attention is paid to those elements of perception and metaphoric interpretation that cause the emotions to rise. Consequences for the design process are then considered, and a design process is proposed that takes into account emotional impacts as one of the goals. A solid scientific approach to the subject is maintained throughout, and understanding is facilitated by the inclusion of a rich collection of successful design artifacts, the emotional aspects of which are analyzed.

  4. Design basis and requirements for 241-SY Modular Exhauster concrete pad and retaining wall

    International Nuclear Information System (INIS)

    Kriskovich, J.R.

    1994-01-01

    The purpose of this document is to serve as the design and functional requirements for a concrete pad for the new 241-SY Modular Exhauster and for a retaining wall to be built near the new ventilation systems

  5. Pupil filter design by using a Bessel functions basis at the image plane.

    Science.gov (United States)

    Canales, Vidal F; Cagigal, Manuel P

    2006-10-30

    Many applications can benefit from the use of pupil filters for controlling the light intensity distribution near the focus of an optical system. Most of the design methods for such filters are based on a second-order expansion of the Point Spread Function (PSF). Here, we present a new procedure for designing radially-symmetric pupil filters. It is more precise than previous procedures as it considers the exact expression of the PSF, expanded as a function of first-order Bessel functions. Furthermore, this new method presents other advantages: the height of the side lobes can be easily controlled, it allows the design of amplitude-only, phase-only or hybrid filters, and the coefficients of the PSF expansion can be directly related to filter parameters. Finally, our procedure allows the design of filters with very different behaviours and optimal performance.

  6. Ergonomic Analysis of Tricycle Sidecar Seats: Basis for Proposed Standard Design

    Directory of Open Access Journals (Sweden)

    Michael C. Godoy

    2015-12-01

    Full Text Available Ergonomics (also called human factors engineering is the study of human characteristics for the appropriate design of the living and work environment. It is applied in various industrial areas which includes transportation.Tricycle being one of the most common means of public transportation in Lipa City has various adaptations to suit the culture, and environment. The purpose of this study is to analyze the variability in design of the tricycles in Lipa City, Philippines and propose a standard ergonomically designed tricycle sidecar seat for a greater population. The study was conducted at 26 tricycle terminals with 232 tricycle samples within Lipa City proper including the public market area where 400 commuters were given questionnaires to determine the risk factors associated with the existing tricycle sidecar seat design. Anthropometric measurements of 100 males and 100 female commuters were obtained together with the sidecar dimensions of 232 tricycles to substantiate the observed variations in design. Using the design for the average and design for the extremes, it was found out that most of the tricycles in Lipa City, Philippines have inappropriate inclined seat and lowered sidecar seat pan height which can result to leg and abdominal pain; narrowed seat pan depth which caused pressure on buttocks and legs; narrowed backrest width which can cause upper and low back pain; low backrest height that can pose upper back pain; which can also result to abdominal pain; inclined backrest and limited vertical clearance which can cause upper back pain and neck pain. The researcher proposed a sidecar seat design standard which can be used by the Land Transportation Office, and Land Transportation Franchising and Regulatory Board to provide ease, comfort, and convenience to the passengers.

  7. Design criteria and design basis for the 100-HR-3 and 100-KR-4 pump-and-treat projects

    International Nuclear Information System (INIS)

    McKinley, W.S.; Winters, J.N.

    1996-06-01

    The 100-HR-3 and 100-KR-4 Operable Units are located in the 100 Area at the Hanford Site in Richland, Washington. The document describes the project objectives and design criteria to be used for the 100-HR-3 and 100-KR-4 groundwater pump-and-treat design activities

  8. Designing Smart Artifacts for Adaptive Mediation of Social Viscosity: Triadic Actor-Network Enactments as a Basis for Interaction Design

    Science.gov (United States)

    Salamanca, Juan

    2012-01-01

    With the advent of ubiquitous computing, interaction design has broadened its object of inquiry into how smart computational artifacts inconspicuously act in people's everyday lives. Although user-centered design approaches remains useful for exploring how people cope with interactive systems, they cannot explain how this new breed of…

  9. Results of the reliability investigations for the design basis accident 'Rupture of a cold primary coolant system'

    International Nuclear Information System (INIS)

    Hoertner, H.; Nieckau, E.; Spindler, H.

    1976-12-01

    This report gives a comprehensive presentation of the detailed reliability investigation carried out for the engineered safety features installed to cope with the design basis accident 'Large LOCA' of a German nuclear power plant with pressurized water reactor. The investigation is based on the engineered safety features of the Biblis Nuclear Power Plant, Unit A. The reliability investigation is carried out by means of a fault tree analysis. The influence of common-mode failures is assessed. (orig.) [de

  10. Experimental study of ballooning and failure of WWER-1000 fuel cans during maximum design basis accident

    International Nuclear Information System (INIS)

    Karetnikov, G.V.; Bogdanov, A.S.; Semishkin, V.P.; Bezrukov, Yu.A.; Trushin, A.M.; Frizen, E.A.

    2001-01-01

    The processes of ballooning and fracturing in tubular specimens of Eh635 and Eh110 alloy fuel cans are investigated with the use of cinematography. The investigations are carried out under steady-state conditions in the temperature range from 680 to 900 deg C and at pressure drops on the can from 2 to 12 MPa. Time dependences of circumferential strains are plotted for various temperatures of fuel cans at pressure of 2 MPa. It is shown that strain changes are of linear character at an initial portion of the curve and then an accelerated strain development takes place with transition to fracture. Using methods of nonlinear evaluation for time to fracture the approximation dependences are obtained for fuel cans. Experimental data are intended to form the equations of state for fuel can materials and to verify the program TVEL-3 [ru

  11. Application of high temperature superconductivity to electric motor design

    International Nuclear Information System (INIS)

    Edmonds, J.S.; Sharma, D.K.; Jordan, H.E.; Edick, J.D.; Schiferl, R.F.

    1992-01-01

    This paper reports on progress made in a joint project conducted by the Electric Power Research Institute and Reliance Electric Company to study the possible application of High Temperature Super Conductors (HTSC), materials to electric motors. Specific applications are identified which can be beneficially served by motors constructed with HTSC materials. A summary is presented of the components and design issues related to HTSC motors designed for these applications. During the course of this development program, a three tier HTSC wire performance specification has evolved. The three specifications and the rationale behind these three levels of performance are explained. A description of a test motor that has been constructed to verify the electromagnetic analytical techniques of HTSC motor design is given. Finally, a DC motor with an HTSC field coil is described. Measured data with the motor running is presented showing that the motor is operating with the field winding in the superconducting state

  12. Engineering design of a high-temperature superconductor current lead

    International Nuclear Information System (INIS)

    Niemann, R.C.; Cha, Y.S.; Hull, J.R.; Daugherty, M.A.; Buckles, W.E.

    1993-01-01

    As part of the US Department of Energy's Superconductivity Pilot Center Program, Argonne National Laboratory and Superconductivity, Inc., are developing high-temperature superconductor (HTS) current leads suitable for application to superconducting magnetic energy storage systems. The principal objective of the development program is to design, construct, and evaluate the performance of HTS current leads suitable for near-term applications. Supporting objectives are to (1) develop performance criteria; (2) develop a detailed design; (3) analyze performance; (4) gain manufacturing experience in the areas of materials and components procurement, fabrication and assembly, quality assurance, and cost; (5) measure performance of critical components and the overall assembly; (6) identify design uncertainties and develop a program for their study; and (7) develop application-acceptance criteria

  13. Design of High Field Solenoids made of High Temperature Superconductors

    Energy Technology Data Exchange (ETDEWEB)

    Bartalesi, Antonio; /Pisa U.

    2010-12-01

    This thesis starts from the analytical mechanical analysis of a superconducting solenoid, loaded by self generated Lorentz forces. Also, a finite element model is proposed and verified with the analytical results. To study the anisotropic behavior of a coil made by layers of superconductor and insulation, a finite element meso-mechanic model is proposed and designed. The resulting material properties are then used in the main solenoid analysis. In parallel, design work is performed as well: an existing Insert Test Facility (ITF) is adapted and structurally verified to support a coil made of YBa{sub 2}Cu{sub 3}O{sub 7}, a High Temperature Superconductor (HTS). Finally, a technological winding process was proposed and the required tooling is designed.

  14. Engineering design of a high-temperature superconductor current lead

    Science.gov (United States)

    Niemann, R. C.; Cha, Y. S.; Hull, J. R.; Daugherty, M. A.; Buckles, W. E.

    As part of the US Department of Energy's Superconductivity Pilot Center Program, Argonne National Laboratory and Superconductivity, Inc., are developing high-temperature superconductor (HTS) current leads suitable for application to superconducting magnetic energy storage systems. The principal objective of the development program is to design, construct, and evaluate the performance of HTS current leads suitable for near-term applications. Supporting objectives are to (1) develop performance criteria; (2) develop a detailed design; (3) analyze performance; (4) gain manufacturing experience in the areas of materials and components procurement, fabrication and assembly, quality assurance, and cost; (5) measure performance of critical components and the overall assembly; (6) identify design uncertainties and develop a program for their study; and (7) develop application-acceptance criteria.

  15. Operation and design selection of high temperature superconducting magnetic bearings

    International Nuclear Information System (INIS)

    Werfel, F N; Floegel-Delor, U; Riedel, T; Rothfeld, R; Wippich, D; Goebel, B

    2004-01-01

    Axial and radial high temperature superconducting (HTS) magnetic bearings are evaluated by their parameters. Journal bearings possess advantages over thrust bearings. High magnetic gradients in a multi-pole permanent magnet (PM) configuration, the surrounding melt textured YBCO stator and adequate designs are the key features for increasing the overall bearing stiffness. The gap distance between rotor and stator determines the specific forces and has a strong impact on the PM rotor design. We report on the designing, building and measuring of a 200 mm prototype 100 kg HTS bearing with an encapsulated and thermally insulated melt textured YBCO ring stator. The encapsulation requires a magnetically large-gap (4-5 mm) operation but reduces the cryogenic effort substantially. The bearing requires 3 l of LN 2 for cooling down, and about 0.2 l LN 2 h -1 under operation. This is a dramatic improvement of the efficiency and in the practical usage of HTS magnetic bearings

  16. Physics basis for an axicell design for the end plugs of MFTF-B

    International Nuclear Information System (INIS)

    Baldwin, D.E.; Logan, B.G.

    1982-01-01

    The primary motivation for conversion of MFTF-B to an axicell configuration lies in its engineering promise as a reactor geometry based on circular high-magnetic-field coils. In comparing this configuration to the previous A-cell geometry, we find a number of differences that might significantly affect the physics performance. The purpose of the present document is to examine those features and to assess their impact on the performance of the axicell, as compared to the A-cell configuration, for MFTF-B. In so doing, we address only those issues thought to be affected by the change in geometry and refer to the original report Physics Basis for MFTF-B, for discussion of those issues thought not to be affected. In Sec. 1, we summarize these physics issues. In Sec. 2, we describe operating scenarios in the new configuration. In the Appendices, we discuss those physics issues that require more detailed treatment

  17. Numerical modeling of the equatorial ionization anomaly (EIA), equatorial temperature and wind anomaly (ETWA) and equatorial electron temperature anomaly (EETA) on the basis of the GSM TIP

    Science.gov (United States)

    Klimenko, M. V.; Klimenko, V. V.; Bryukhanov, V. V.

    On the basis of Global Self-consistent Model of Thermosphere Ionosphere and Protonosphere GSM TIP developed in WD IZMIRAN the calculations of the behavior of thermosphere F-region and upper ionosphere parameters at middle and low geomagnetic latitudes are carried out The calculations were carried out with use the new block of the calculation of electric fields in the ionosphere in which the decision of the three-dimensional equation describing the law of the conservation of the full current density in the ionosphere of the Earth is realized by adduction it to the two-dimensional by integration on the thickness of the current conductive layer of the ionosphere along equipotential geomagnetic field lines The calculations of the neutral atmosphere composition and temperature were executed with use of the MSIS model The quite geomagnetic conditions of the equinox were considered in the minimum of the solar activity There are presented the calculated global distributions of the critical frequency of the F2-layer of ionosphere for the different moments UT the latitudinal course of the N e and T e in the F-region and upper ionosphere in the vicinity of geomagnetic equator and unrolling on UT of the calculated velocities of zonal component of the thermospheric wind and ion temperature in the F-region of ionosphere as well as critical frequency and height of the F2-layer maximum of the ionosphere at three longitude chains of the stations Brazilian -- Fortaleza 4 0 r S 38 0 r W Jicamarca 11 9 r S 76 0 r W Cachoeira

  18. Microstructural design of magnesium alloys for elevated temperature performance

    Science.gov (United States)

    Bryan, Zachary Lee

    Magnesium alloys are promising for automotive and aerospace applications requiring lightweight structural metals due to their high specific strength. Weight reductions through material substitution significantly improve fuel efficiency and reduce greenhouse gas emissions. Challenges to widespread integration of Mg alloys primarily result from their limited ductility and elevated temperature strength. This research presents a microstructurally-driven systems design approach to Mg alloy development for elevated temperature applications. The alloy properties that were targeted included creep resistance, elevated temperature strength, room temperature ductility, and material cost. To enable microstructural predictions during the design process, computational thermodynamics was utilized with a newly developed atomic mobility database for HCP-Mg. The mobilities for Mg self-diffusion, as well as Al, Ag, Sn, and Zn solute diffusion in HCP-Mg were optimized from available diffusion literature using DICTRA. The optimized mobility database was then validated using experimental diffusion couples. To limit dislocation creep mechanisms in the first design iteration, a microstructure consisting of Al solutes in solid solution and a fine dispersion of Mg2Sn precipitates was targeted. The development of strength and diffusion models informed by thermodynamic predictions of phase equilibria led to the selection of an optimum Mg-1.9at%Sn-1.5at%Al (TA) alloy for elevated temperature performance. This alloy was cast, solution treated based upon DICTRA homogenization simulations, and then aged. While the tensile and creep properties were competitive with conventional Mg alloys, the TA mechanical performance was ultimately limited because of abnormal grain growth that occurred during solution treatment and the basal Mg2Sn particle morphology. For the second design iteration, insoluble Mg2Si intermetallic particles were added to the TA alloy to provide enhanced grain boundary pinning

  19. Planning of designing and installation of mechanical elements at the gear speed reducer on the basis of the parameter technology

    Directory of Open Access Journals (Sweden)

    D. Letić

    2013-01-01

    Full Text Available The development and implementation of the computer methods at project managing in the part of the planning of designing and installation of mechanical elements with the fit (assembly block of the gear speed reducer is significant and at present irreplaceable engineering task if it has been realized by the modern parameter technology. There are multifunction uses of this organized group of activities, beginning from the quick changeability of elements still in the phase of designing and constructing, thanks to the characteristics of their associativity, still to the wide basis of standard elements that are incorporated in the very program package. Meanwhile, these activities are not simple, so their realization has to be planned from the stand - point of time, resource and cost of realization. For the very designing and constructing was used AutoCAD Mechanical, and for the design managing Microsoft Project.

  20. Study for optimizing the design of optical temperature sensor

    Science.gov (United States)

    Li, Panpan; Sun, Zhen; Shi, Ruixin; Liu, Guofeng; Fu, Zuoling; Wei, Yanling

    2017-12-01

    The correlations between temperature sensitivity (relative sensitivity Sr and absolute sensitivity Sa) and thermally coupled level gaps (ΔE) are vital but less-studied for potential applications in scientific research, industrial production, clinical medicine, and so on. We take YbPO4:Ln3+ (Ln = Tm3+, Ho3+, and Er3+) up-conversion phosphors as a case to study the relationships between temperature sensitivity (Sr, Sa) and ΔE. The results of various discussions, including the experimental data of temperature sensitivity based on YbPO4:Ln3+ (Ln = Tm3+, Ho3+, and Er3+) and theoretical derivation from original formulas, show that Sr and ΔE are linearly positive correlation, which is invalid for Sa. Noticeably, YbPO4:Tm3+ nanoparticles display intense near infrared red emission within the biological window, leading to great potential application in biological sensing and biological imaging. All the research studies would benefit the design of optical temperature sensing.

  1. Design of high temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

    1994-09-01

    Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

  2. Groundwork: Preparing an Effective Basis for Communication and Shared Learning in Design and Technology Education

    Science.gov (United States)

    Looijenga, Annemarie; Klapwijk, Remke; de Vries, Marc J.

    2016-01-01

    In Dutch Design and Technology Education the beginning of a process of learning is usually determined by the teacher. In this paper it is argued that a beginning, determined in interaction with the students, is more profitable as the interaction will lead to joined-up exploring, creating and thinking and an increased motivation to learn.…

  3. Workshops for professionals as basis for students' workshops for sustainable design

    NARCIS (Netherlands)

    Zeiler, W.; Houten, van M.A.; Savanovic, P.; Kim, S.; Chen, L.

    2009-01-01

    The growing complexity due to the increased demand for more sustainability in (Dutch) building practice necessitates developments in other aspects, besides specialized and professional skills. Therefore a new integral approach in building design education has been developed in close cooperation with

  4. Design and application consideration of high temperature superconducting current leads

    International Nuclear Information System (INIS)

    Wu, J.L.

    1994-01-01

    As a potential major source of heat leak and the resultant cryogen boiloff, cryogenic current leads can significantly affect the refrigeration power requirement of cryogenic power equipment. Reduction of the heat leak associated with current leads can therefore contribute to the development and application of this equipment. Recent studies and tests have demonstrated that, due to their superconducting and low thermal conductivity properties, ceramic high temperature superconductor (HTSC) can be employed in current leads to significantly reduce the heat leak. However, realization of this benefit requires special design considerations pertaining to the properties and the fabrication technology of the relatively new ceramic superconductor materials. Since processing and fabrication technology are continuously being developed in the laboratories, data on material properties unrelated to critical states are quite limited. Therefore, design analysis and experiments have to be conducted in tandem to achieve a successful development. Due to the rather unique combination of superconducting and thermal conductivities which are orders of magnitude lower than copper, ceramic superconductors allow expansion of the operating scenarios of current leads. In addition to the conventional vapor-cooled lead type application, low heat leak conduction-cooled type current leads may be practical and are being developed. Furthermore, a current lead with an intermediate heat leak intercept has been successfully demonstrated in a multiple current lead assembly employing HTSC. These design and application considerations of high temperature superconducting current leads are addressed here

  5. Design and Control of High Temperature PEM Fuel Cell System

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl

    E-cient fuel cell systems have started to appear in many dierent commercial applications and large scale production facilities are already operating to supply fuel cells to support an ever growing market. Fuel cells are typically considered to replace leadacid batteries in applications where...... to conventional PEM fuel cells, that use liquid water as a proton conductor and thus operate at temperatures below 100oC. The HTPEM fuel cell membrane in focus in this work is the BASF Celtec-P polybenzimidazole (PBI) membrane that uses phosphoric acid as a proton conductor. The absence of water in the fuel cells...... enables the use of designing cathode air cooled stacks greatly simplifying the fuel cell system and lowering the parasitic losses. Furthermore, the fuel impurity tolerance is signicantly improved because of the higher temperatures, and much higher concentrations of CO can be endured without performance...

  6. Comparison of analyzed design-basis events to actual plant transients

    International Nuclear Information System (INIS)

    Geeting, M.W.; Hightower, N.T. III; Fields, C.C.

    1992-01-01

    Fitness-for-Service Guidelines have recently been developed to provide acceptance criteria and evaluation methods for assessment of the integrity of the Zr-2.5 Nb pressure tubes in operating Canada deuterium uranium (CANDU) reactors. The guidelines provide a methodology for the evaluation of specific conditions in a single tube, such as manufacturing and inservice generated flaws, hydride blisters formed at points of contact between a pressure tube and its calandria tube, and generic degradation of pressure tube properties in service. The guidelines are divided into three sections. The first section describes the requirements that must be met to qualify the tubes ofr continued service. The second section provides the material properties data-base information needed to carry out the assessments. The third section provides the technical basis for the acceptance criteria and evaluation procedures as well as justifications and descriptions of the data bases. The guidelines were issued to CANDU reactor operators for trial use and released to the Atomic Energy Control Board of Canada for review and comment in May 1991

  7. Environment as a basis for the design of advertising structures by forming

    Science.gov (United States)

    Khmelevsky, Y. P.; Seryakov, V. A.; Mamontov, G. Y.; Tsarenko, D. T.

    2017-01-01

    A few different neighbouring styles of architectural forms are quite frequent in the cities of great historical past. As a result, a designer or architect has to solve the complex problem while designing the objects within such environment, i.e. one has to fit them naturally into the existing site development. Often, form making is found to be hard, due to the fact that the existing architectural forms of totally different stylistic execution coexist in the visual proximity. Presently, placement of the advertising bills in urban environment is both an urgent and debatable issue. On the one hand, advertising providers are keen to present their product bigger and brighter, on the other hand, the overall and eye-catching exhibition stands can be disharmonious with the surrounding architectural ensemble of the city. This situation is relevant for every cultural city.

  8. Design and Modelling of Small Scale Low Temperature Power Cycles

    DEFF Research Database (Denmark)

    Wronski, Jorrit

    he work presented in this report contributes to the state of the art within design and modelling of small scale low temperature power cycles. The study is divided into three main parts: (i) fluid property evaluation, (ii) expansion device investigations and (iii) heat exchanger performance......-oriented Modelica code and was included in the thermo Cycle framework for small scale ORC systems. Special attention was paid to the valve system and a control method for variable expansion ratios was introduced based on a cogeneration scenario. Admission control based on evaporator and condenser conditions...

  9. Conceptual designs for advanced, high-temperature CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushby, S.J. [Atomic Energy of Canada Ltd., Corrosion and Surface Science Branch, Chalk River Laboratories, Chalk River, ON (Canada); Dimmick, G.R. [Atomic Energy of Canada Ltd., Fuel Channel Thermmalhydraulics Branch, Chalk River, ON (Canada); Duffey, R.B. [Atomic Energy of Canada Ltd., Principal Scientist, Chalk River Laboratories, Chalk River, On (Canada); Spinks, N.J. [Atomic Energy of Canada Ltd., Researcher Emeritus, Chalk River Laboratories, Chalk River, ON (Canada); Burrill, K.A. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, ON (Canada); Chan, P.S.W. [Atomic Energy of Canada Ltd., Reactor Core Physics Branch, Mississauga, ON (Canada)

    2000-07-01

    AECL is studying advanced reactor concepts with the aim of significant cost reduction through improved thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, also incorporates enhanced safety features, and flexible, proliferation-resistant fuel cycles, whilst retaining the fundamental design characteristics of CANDU: neutron economy, horizontal fuel channels, and a separate D{sub 2}O moderator that provides a passive heat sink. Where possible, proven, existing components and materials would be adopted, so that 'first-of-a-kind' costs and uncertainties are avoided. Three reactor concepts ranging in output from {approx}375 MW(e) to 1150 MW(e) are described. The modular design of a pressure tube reactor allows the plant size for each concept to be tailored to a given market through the addition or removal of fuel channels. Each concept uses supercritical water as the coolant at a nominal pressure of 25 MPa. Core outlet temperatures range from {approx}400degC to 625degC, resulting in substantial improvements in thermodynamic efficiencies compared to current nuclear stations. The CANDU-X Mark 1 concept is an extension of the present CANDU design. An indirect cycle is employed, but efficiency is increased due to higher coolant temperature, and changes to the secondary side; as well, the size and number of pumps and steam generators are reduced. Safety is enhanced through facilitation of thermo-siphoning of decay heat by increasing the temperature of the moderator. The CANDU-X NC concept is also based on an indirect cycle, but natural convection is used to circulate the primary coolant. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the pseudo-critical temperature of water because of large changes in heat capacity and thermal expansion in that region. In the third concept (CANDUal-X), a dual cycle is employed. Supercritical water exits the core and feeds directly into a very high

  10. Beyond design basis external flooding. Generic design assessment and lessons learned from the Fukushima Dai-ichi accident

    International Nuclear Information System (INIS)

    MacLeod, Tanya; Smith, Leslie; Allmark, Tim; Ford, Peter

    2017-01-01

    New reactors intended for construction in GB undergo the Office for Nuclear Regulation's (ONR's) Generic Design Assessment (GDA). GDA is a pre-licensing process that provides requesting parties with the opportunity to demonstrate at an early stage that the design is capable of meeting the legal requirements of Great Britain. During GDA, the intended reactor site may not yet be known. Therefore, requesting parties usually define a 'Generic Site' with characteristics typical for Great Britain. These characteristics should, as far as possible, bound the characteristics of known potential sites so that reactors of the proposed type could potentially be built at various suitable locations. This paper critically reviews ONR's approach to ensuring that external flooding is appropriately addressed at the GDA stage and covers: An overview of ONR's approach to post-Fukushima assessment. Changes to ONR's SAPs (Safety Assessment Principles) related to External Flooding. Two examples of post-Fukushima GDA approaches to External Flooding. Uncertainty and the provision of adequate safety measures. The paper concludes that the identification of potential vulnerabilities in the design to external flooding combined with a consideration of post-Fukushima resilience enhancements has led to increased regulatory confidence in the robustness of new reactor designs in GB against external flooding. (author)

  11. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-01-01

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. Therefore, empirically based approaches that are used for other regions, such as Western North America, are not appropriate for Eastern North America. Moreover, recent advances in science and technology have now made it possible to combine theoretical and empirical methods to develop new procedures and models for estimating ground motion. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. Specifically considered are magnitudes M from 5 to 8, distances from 0 to 500 km, and frequencies from 1 to 35 Hz. This document, Volume IV, provides Appendix 8.B, Laboratory Investigations of Dynamic Properties of Reference Sites

  12. Design basis and requirements for 241-SY modular exhauster mechanical installation

    International Nuclear Information System (INIS)

    Kriskovich, J.R.

    1994-01-01

    A new ventilation system is being installed to serve as the K-1 primary exhauster. The existing K-1 primary exhauster will then become the backup. This ventilation system services waste tanks 241-SY-101, 102 and 103. The nominal flow rate through the ventilation system is 1,000 cfm. The new ventilation system will contain a moisture eliminator, a heater, a prefilter, two stages of HEPA filtration, an exhaust fan, a stack and stack sampling system. The purpose of this document is to serve as the design and functional requirements for the mechanical installation of the new 241-SY modular exhauster. The mechanical installation will include modifying the existing ductwork (i.e., installing a ''T'' to connect the new exhauster to the existing system), modifying the existing condensate drain lines to accommodate the new lines associated with the new exhauster, a maintenance platform near the stack of the new exhauster, guy wires and guy wire footings to support the stack of the new exhauster, as well as other miscellaneous tasks associated with the mechanical installation design effort

  13. Protein structures in Alzheimer's disease: The basis for rationale therapeutic design.

    Science.gov (United States)

    Montoliu-Gaya, Laia; Villegas, Sandra

    2015-12-15

    Alzheimer's disease (AD) is a neurodegenerative disorder that affects memory, behavior, thinking and emotion. Current therapies to treat AD patients are only capable for temporarily slowing-down the cognitive decline, as they are focused on ameliorating symptoms instead of targeting its underlying causes. The aim of this review is to describe what is known about the protein structures implicated in AD pathogenesis, amyloid cascade members, as well as those structures involved in Aβ clearance. Thus, structural information available for APP, α- β- and γ-secretases, CTFβ and derived Aβ peptides, AICDs, apoE and apoJ, LRP-1 and RAGE, and neprilysin and insulin-degrading enzyme is provided. The recently solved structure for the γ-secretase complex opens the rational design of a new generation of inhibitors, whereas that for Aβ oligomers offers a putative mechanism explaining why monoclonal antibodies targeted to the N-terminus are effective. Then, an overview on therapies targeting some of these molecules presents their benefits and drawbacks. As a general conclusion our knowledge on the protein structures involved in AD has recently substantially advanced, allowing for the rational design of different therapeutic approaches. Hopefully, we are getting closer to finding a strong disease-modifying drug to cure this devastating disease. Copyright © 2015 Elsevier Inc. All rights reserved.

  14. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-01-01

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. Therefore, empirically based approaches that are used for other regions, such as Western North America, are not appropriate for Eastern North America. Moreover, recent advances in science and technology have now made it possible to combine theoretical and empirical methods to develop new procedures and models for estimating ground motion. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. Specifically considered are magnitudes M from 5 to 8, distances from 0 to 500 km, and frequencies from 1 to 35 Hz

  15. Cost analysis of a commercial pyroprocess facility on the basis of a conceptual design in Korea

    International Nuclear Information System (INIS)

    Kim, S.K.; Ko, W.I.; Youn, S.R.; Gao, Ruxing

    2015-01-01

    Highlights: • Pyroprocess facility’s direct cost was calculated based on the conceptual design. • The unit cost of pyroprocess was calculated as $781/kgHM. • The unit cost was increased by 3%, considering labor allocation standards. • The operating and maintenance cost was identified as a main cost driver. - Abstract: This study postulated a commercial pyroprocess facility (KAPF+: Korea Advanced Pyroprocess Facility Plus) with a processing capacity of 400 tons/year as a cost object, and utilized an engineering cost estimation method based on a conceptual design to present the results of the total cost and unit cost estimation. According to the calculation results, the total cost and unit cost were calculated with k$779,386 and $781/kgHM, respectively. Moreover, the key cost driver was manifested as the operating and maintenance costs. In particular, equipment replacement cost was identified as an important cost driver. In addition, for an increasingly accurate cost estimation, the calculation results and allocation method of the indirect cost were reanalyzed. Finally the pyroprocess unit cost increased $5 when calculated the indirect cost using the labor time as the allocation standard. Meanwhile, the pyroprocess unit cost increased $22 as a result of allocating the indirect cost using the uniform labor cost as the cost allocation standard. Accordingly, an indirect cost allocation standard was manifested as the factor that exerts a significant effect on the pyroprocess unit cost

  16. Status of design code work for metallic high temperature components

    International Nuclear Information System (INIS)

    Bieniussa, K.; Seehafer, H.J.; Over, H.H.; Hughes, P.

    1984-01-01

    The mechanical components of high temperature gas-cooled reactors, HTGR, are exposed to temperatures up to about 1000 deg. C and this in a more or less corrosive gas environment. Under these conditions metallic structural materials show a time-dependent structural behavior. Furthermore changes in the structure of the material and loss of material in the surface can result. The structural material of the components will be stressed originating from load-controlled quantities, for example pressure or dead weight, and/or deformation-controlled quantities, for example thermal expansion or temperature distribution, and thus it can suffer rowing permanent strains and deformations and an exhaustion of the material (damage) both followed by failure. To avoid a failure of the components the design requires the consideration of the following structural failure modes: ductile rupture due to short-term loadings; creep rupture due to long-term loadings; reep-fatigue failure due to cyclic loadings excessive strains due to incremental deformation or creep ratcheting; loss of function due to excessive deformations; loss of stability due to short-term loadings; loss of stability due to long-term loadings; environmentally caused material failure (excessive corrosion); fast fracture due to instable crack growth

  17. Conceptual designs for very high-temperature CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

    2000-07-01

    Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310{sup o}C, and exits at {approx}410{sup o}C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed

  18. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B.

    2000-01-01

    Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310 o C, and exits at ∼410 o C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed power plants

  19. Design and development of self-powered sensors on wireless sensor network for standalone plant critical data management during SBO and beyond design basis events

    International Nuclear Information System (INIS)

    Aparna, J.; Dulera, I.V.; Rama Rao, A.; Vijayan, P.K.

    2015-01-01

    Advanced reactors are designed with an aim of maximum safety, optimized fuel utilization and effective system design. Safety aspects in reactor designs are being viewed for all possible vulnerabilities, and as a result, robust self-regulating passive safety features have been favored in Gen IV and advanced reactor designs. In addition to passive systems, the accidents scenarios at Fukushima indicate the dire need of reliable and stand-alone self-powered sensors, for monitoring plant critical parameters for effective damage control actions. There is a strong need for plant critical data management and situation awareness during the unavailability of all conventional power sources in a nuclear power plant, during extended station blackout (SBO) conditions. These self-powered sensors would assist the operators in managing events like SBO and help in containing any Beyond Design Basis Events (BDBE) conditions, well away from the public domain

  20. Basis for the safety approach for design and assessment of Generation IV nuclear systems

    International Nuclear Information System (INIS)

    Fiorini, G.L.; Leahy, T.

    2009-01-01

    The primary objective of the RSWG is the implementation of a harmonized approach on long-term safety, and to address risk and regulatory issues in development of the next generation of nuclear systems. To this end, the group is proposing safety goals and evaluation methodology applicable for the design and assessment of future systems. The paper resumes the content of the first RSWG report which provides insights for the safety approach and assists the GIF Systems Steering Committee as well as the GIF Experts Group and the GIF Policy Group for the definition of the most adequate safety related Gen IV R and D. The document is also an essential contributor to help identifying the needed supportive crosscut R and D effort (i.e. applicable to all the innovative nuclear technologies). Although the report presents a number of thoughts and recommendations, it really represents only the start of the efforts for the RSWG. (author)

  1. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    International Nuclear Information System (INIS)

    Ross, Kyle; Cardoni, Jeffrey N.; Wilson, Chisom Shawn; Morrow, Charles; Osborn, Douglas; Gauntt, Randall O.

    2015-01-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  2. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, Chisom Shawn [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, Douglas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  3. TRAC analysis of design basis events for the accelerator production of tritium target/blanket

    International Nuclear Information System (INIS)

    Lin, J.C.; Elson, J.

    1997-01-01

    A two-loop primary cooling system with a residual heat removal system was designed to mitigate the heat generated in the tungsten neutron source rods inside the rungs of the ladders and the shell of the rungs. The Transient Reactor Analysis Code (TRAC) was used to analyze the thermal-hydraulic behavior of the primary cooling system during a pump coastdown transient; a cold-leg, large-break loss-of-coolant accident (LBLOCA); a hot-leg LBLOCA; and a target downcomer LBLOCA. The TRAC analysis results showed that the heat generated in the tungsten neutron source rods can be mitigated by the primary cooling system for the pump coastdown transient and all the LBLOCAs except the target downcomer LBLOCA. For the target downcomer LBLOCA, a cavity flood system is required to fill the cavity with water at a level above the large fixed headers

  4. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-01-01

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. This document, Volume II, contains Appendices 2, 3, 5, 6, and 7 covering the following topics: Eastern North American Empirical Ground Motion Data; Examination of Variance of Seismographic Network Data; Soil Amplification and Vertical-to-Horizontal Ratios from Analysis of Strong Motion Data From Active Tectonic Regions; Revision and Calibration of Ou and Herrmann Method; Generalized Ray Procedure for Modeling Ground Motion Attenuation; Crustal Models for Velocity Regionalization; Depth Distribution Models; Development of Generic Site Effects Model; Validation and Comparison of One-Dimensional Site Response Methodologies; Plots of Amplification Factors; Assessment of Coupling Between Vertical ampersand Horizontal Motions in Nonlinear Site Response Analysis; and Modeling of Dynamic Soil Properties

  5. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions. Volume 2, Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1993-03-18

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake`s ground motion is a function of the earthquake`s magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. This document, Volume II, contains Appendices 2, 3, 5, 6, and 7 covering the following topics: Eastern North American Empirical Ground Motion Data; Examination of Variance of Seismographic Network Data; Soil Amplification and Vertical-to-Horizontal Ratios from Analysis of Strong Motion Data From Active Tectonic Regions; Revision and Calibration of Ou and Herrmann Method; Generalized Ray Procedure for Modeling Ground Motion Attenuation; Crustal Models for Velocity Regionalization; Depth Distribution Models; Development of Generic Site Effects Model; Validation and Comparison of One-Dimensional Site Response Methodologies; Plots of Amplification Factors; Assessment of Coupling Between Vertical & Horizontal Motions in Nonlinear Site Response Analysis; and Modeling of Dynamic Soil Properties.

  6. Design basis for the copper/steel canister. Stage three. Final report

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    1997-02-01

    The development of the copper/iron canister proposed for the containment of high-level waste in the Swedish disposal programme has been studied from the points of view of choice of materials, manufacturing technology and Q A. This report describes the observations on progress which has been made between March 1995 and February 1996 and the results of further literature studies. A first trial canister has been produced by SKB using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. It is considered that such a change will require a significant development programme. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. An improved microstructure may be achieved by extruding at a lower temperature but this remains to be demonstrated. Similar problems exist with plate used for the fabricated tubular but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. However it was necessary to constrain it during welding and it subsequently distorted during machining. There was some evidence of hot tearing close to the weld. The distortion problem may be overcome by a stress relieving anneal but this could cause further grain size problems. 19 refs

  7. Romanian experience and approach in developing the design basis threat concept

    International Nuclear Information System (INIS)

    Ionescu, D.S.

    2002-01-01

    Full text: Physical protection systems required to protect the nuclear facilities and materials has to be able to respond at least to the design base threat (DBT). First, the DBT is defined as a minimum set of characteristics of a probable adversary in order to give to the security designer the necessary information to design the Physical Protection System (PPS) subsystems at least at the level to assure complete protection of the nuclear material and facilities against certain attacks. The secondary role of a DBT is to evaluate the effectiveness of an already built PPS facing with a threat having the general characteristic described by DBT. Romanian process to define the natural DBT started effectively with a workshop organized by AIEA and the Romanian Nuclear Regulatory Commission (CNCAN). There participated all the Romanian institution involved in the field of protection of nuclear material and facilities (Romanian Intelligence, Military, Police, CNCAN) as well the representatives of main nuclear operators. The workshop scope was to emphasize the methodology to build a DBT instead to define that. The participants were able to understand the necessary steps to define DBT and the importance of the teamwork. After that meeting, under CNCAN coordination started a process to achieve information from the state institution and nuclear operators. Various meeting has been held to refine the information achieved. Those meetings constituted an important step in the process, helping the parts to understand each other (before that, due to the different organizations where the participants come from was difficult to achieve a common language in this respect). Lots of specific or general analyses and materials have been evaluated, refined and integrated in a special report. Two main constrains of DBT has been identified: the DBT must be 'strong' enough to permit a powerful response of the PPS to any credible attack; also the DBT should consider the equilibrium between the

  8. GaAs/AlAs/InGaP heterostructure: a versatile material basis for cantilever designs

    International Nuclear Information System (INIS)

    Gregušová, Dagmar; Kúdela, Róbert; Eliáš, Peter; Šoltýs, Ján; Cambel, Vladimír; Kostič, Ivan

    2010-01-01

    We report on the design, fabrication and initial mechanical testing of cantilevers with tips based on a GaAs/In 0.485 Ga 0.515 P/AlAs heterostructure grown by metal organic chemical vapor deposition. They were produced using a dedicated technological process based on (1) the formation of integrated tips through an AlAs-assisted surface sacrificial wet-etching process and (2) the GaAs cantilever release fully protected between two InGaP etch-stop layers. 2 µm thick InGaP/GaAs/InGaP cantilevers had integrated pyramidal tips with the sides at ∼45° to (1 0 0). Metallic elements were processed close to the tip apexes using non-standard optical lithography. The cantilever release was accomplished using photolithography, Ar ion milling of InGaP and wet chemical etching of GaAs via resist layers deposited by a draping technique. A tip–cantilever prototype with length, width and thickness of 150, 35 and 2 µm, respectively, exhibited a resonance frequency of 66.2 kHz, which correlated well with a theoretical value of 57 kHz for a GaAs cantilever of identical dimensions. (technical note)

  9. A comparison of the consequences of the design basis accident of the Greek Research Reactor with those of a serious realistic accident

    International Nuclear Information System (INIS)

    Kollas, J.G.; Anoussis, J.N.

    1985-12-01

    An analysis of the radiological consequences of the design basis and the coolant flow blockage accidents of the Greek Research Reactor is presented. The results indicate that the consequences of the coolant flow blockage accident are practically trivial being 1-2 orders of magnitude lower than the corresponding consequences of the design basis accident. (author)

  10. Session 1 theme: Various forms of design basis knowledge and effects of its loss on Safety. Views from EDF

    International Nuclear Information System (INIS)

    Servière, Georges

    2013-01-01

    Design basis knowledge - What happens or may happen and corresponding required knowledge: • Unexpected events or failures of equipment; • Spare part issues (no longer availaible,…); • Change in applicable regulations / requirements; • Change of operating conditions; • Change of plant performances; • Evolution of external environment and conditions; • Events and accidents on other plants, worldwide; • New knowledge availaible; • Periodic safety reviews and upgrades; • Extension of plant operation life; • Decommissioning and dismantling; • Some of those you may choose not to do, but most of them have to be faced and need appropriate knowledge

  11. Design basis for the copper/steel canister. Stage four. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, W.H. [Meadow End Farm, Farnham (United Kingdom)

    1998-06-01

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the points of view of choice of materials, manufacturing technology and quality assurance. Cast steel has been rejected in favour of cast iron as a candidate material for the load bearing liner. Nodular (or ductile) iron is selected and this is capable of providing mechanical properties which are equally suitable as those of the originally selected high strength low alloy steel. The material specified for the overpack is Oxygen free copper with 50 ppm of phosphorus added. Corrosion studies supported by SKB indicate that in the absence of mechanical failure or accelerated localised corrosion the overpack should provide corrosion shielding of the canister for its full design life. Published work claiming that the nodular iron liner would have corrosion characteristics similar to the carbon steel which had been examined in depth is flawed since the microstructures of the iron and carbon steel specimens used were not investigated. It is highly unlikely that nodular irons in the form used for the experiments would have similar structures to nodular iron in the canisters by chance. If the overpack were breached during the aerobic period of the repository life then very rapid penetration of the inner liner could occur. It has been recognised that the roll forming method is not suitable for serial production and alternatives are being sought. The electron beam welding process has been explored with tenacity but has so far failed to produce a satisfactory lid weld. A new welder is being developed for supply to the SKB pilot plant where development will be continued. An alternative welding process, friction stir welding, is being examined as a candidate for attaching lids. Surface breaking defects may be detected using eddy current methods but there is currently no reliable way of detecting small sub surface defects in the overpack

  12. Design of dynamic loading support on high temperature pipe

    International Nuclear Information System (INIS)

    Sitandung, Y.B.; Bandriyana, B.

    2002-01-01

    As a follow up to pipe stress analysis result caused by high temperature operation loading, a design of dynamic loading support was made. The type of variable and constant support as acceptable choosing are applicated for reduce of over stress and over load on piping system. Analysis line schedule of AP600 as an example with apply three dynamic loading support (two type variable and one type constant support). The pre-design of the third support above are based on analysis result with follow the support catalog and field condition wherein its supports are installed. To guarantee the performance and accurate of the support, checking is performed for spring working rate tolerance, support variability and swing angle. The design results of variable spring are loads, size, working rate, type tolerance, spring rate, variability, long and sway angle with each values 5000; 15; 1,25; VM; 0.655; 1080; 0.114; 114,5; 0,48 for S1 and 2045; 12; 0,583; VS; 0,237; 900; 0,132; 130; 0,34 for S3

  13. Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

    International Nuclear Information System (INIS)

    Heard, F.J.

    1997-01-01

    The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

  14. Inter-regional Knowledge Management Workshop on Life-cycle Management of Design Basis Information – Issues, Challenges, Approaches. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    This Workshop had a strategic focus on identifying and clarifying long-term issues and objectives related to our collective responsibilities to ensure that both existing nuclear facilities and future new build projects properly address life-cycle management of plant design basis knowledge (i.e. from design to decommissioning). The workshop attempted to bring together key stakeholders and build a better collective understanding, recognizing that very different perspectives exist and there are a wide range of national contexts and approaches. The various issues and challenges related to this topic and facing the nuclear energy sector both today and in the long-term were discussed in a senior management context and at strategic level

  15. PCDP [Prototypical Spent Fuel Consolidation Equipment Demonstration Project] design basis accident report 9315-P-103, Rev. A

    International Nuclear Information System (INIS)

    1987-12-01

    The Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) has identified a requirement to integrate the spent fuel rod consolidation design activities of each of several proposed geological repository facilities and the Monitored Retrievable Storage (MRS) facility, and to develop efficient and cost-effective equipment for the consolidation process. The equipment to be developed for the rod consolidation system will be required to operate in a dry environment at rates which can be appropriately scaled to approximate the waste management system acceptance rates, irrespective of repository geologic characteristics or the existence of an MRS facility in the waste management system. The purpose of this report is to identify and analyze the range of facility credible events and accident occurrences (from minor to the design basis accidents) and their causes and consequences. For each situation, the considerations to prevent or mitigate the event or accident is addressed

  16. Design prospect of remountable high-temperature superconducting magnet

    Energy Technology Data Exchange (ETDEWEB)

    Hashizume, Hidetoshi, E-mail: hidetoshi.hashizume@qse.tohoku.ac.jp; Ito, Satoshi

    2014-10-15

    The remountable (mountable and demountable repeatedly) high-temperature superconducting (HTS) magnet has been proposed for huge and complex superconducting magnets in future fusion reactors to fabricate and repair easily the magnet and access inner structural components. This paper summarizes progress in R and D activities of mechanical joints of HTS conductors in terms of the electrical resistance and heat transfer performance at the joint region. The latest experimental results show the low joint resistance, 4 nΩ under 70 kA current condition using REBCO HTS conductor with mechanical lap joint system, and for the cooling system the maximum heat flux of 0.4 MW/m{sup 2} is removed by using bronze sintered porous media with sub-cooled liquid nitrogen. These values indicate that there is large possibility to design the remountable HTS magnet for fusion reactors.

  17. Design of stirling engine operating at low temperature difference

    Directory of Open Access Journals (Sweden)

    Sedlák Josef

    2018-01-01

    Full Text Available There are many sources of free energy available in the form of heat that is often simply wasted. The aim of this paper is to design and build a low temperature differential Stirling engine that would be powered exclusively from heat sources such as waste hot water or focused solar rays. A prototype is limited to a low temperature differential modification because of a choice of ABSplus plastic as a construction material for its key parts. The paper is divided into two parts. The first part covers a brief history of Stirling engine and its applications nowadays. Moreover, it describes basic principles of its operation that are supplemented by thermodynamic relations. Furthermore, an analysis of applied Fused Deposition Modelling has been done since the parts with more complex geometry had been manufactured using this additive technology. The second (experimental part covers 4 essential steps of a rapid prototyping method - Computer Aided Design of the 3D model of Stirling engine using parametric modeller Autodesk Inventor, production of its components using 3D printer uPrint, assembly and final testing. Special attention was devoted to last two steps of the process since the surfaces of the printed parts were sandpapered and sprayed. Parts, where an ABS plus plastic would have impeded the correct function, had been manufactured from aluminium and brass by cutting operations. Remaining parts had been bought in a hardware store as it would be uneconomical and unreasonable to manufacture them. Last two chapters of the paper describe final testing, mention the problems that appeared during its production and propose new approaches that could be used in the future to improve the project.

  18. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

    2014-10-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  19. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    International Nuclear Information System (INIS)

    Lazaro, A.; Schikorr, M.; Mikityuk, K.; Ammirabile, L.; Bandini, G.; Darmet, G.; Schmitt, D.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D.; Stempniewicz, M.

    2014-01-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs

  20. Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

    2015-08-15

    Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

  1. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Maine Yankee nuclear power plant

    International Nuclear Information System (INIS)

    Latorre, V.R.; Mayn, B.G.

    1979-08-01

    This report documents the technical evaluation of the electrical, instrumentation, and control design aspects for the low temperature overpressure protection system of the Maine Yankee nuclear power plant. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria

  2. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Yankee Rowe nuclear power plant

    International Nuclear Information System (INIS)

    Latorre, V.R.; Mayn, B.G.

    1979-08-01

    This report documents the technical evaluation of the electrical, instrumentation, and control design aspects for the low temperature overpressure protection system of the Yankee Rowe nuclear power plant. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria

  3. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Salem nuclear power plant, Unit 1

    International Nuclear Information System (INIS)

    Laudenbach, D.H.

    1979-03-01

    The technical evaluation is presented for the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Salem nuclear power plant, Unit 1. Design basis criteria used to evaluate the acceptability of the system include operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria

  4. The influence of wind on the temperature of PV modules in tropical environments, evaluated on an hourly basis

    NARCIS (Netherlands)

    Veldhuis, A.J.; Nobre, A.; Reindl, T.; Ruther, R.; Reinders, Angelina H.M.E.

    2013-01-01

    It is well known that the efficiency of PV modules decreases with increasing module temperatures. Many studies have paid attention to the development and validation of heuristic models to calculate the PV module temperature in higher latitudes, however only a few focus on the thermal behaviour of PV

  5. Design and fabrication of sintered Nd-Fe-B magnets with a low temperature coefficient of intrinsic coercivity

    Directory of Open Access Journals (Sweden)

    Cui X.G.

    2009-01-01

    Full Text Available To decrease the temperature coefficients of sintered Nd-Fe-B magnets, the influencing factors on temperature coefficients, especially the reversible temperature coefficient β of intrinsic coercivity Hcj, were analyzed. The results showed that the absolute value of β decreased with increasing Hcj and also the ratio of microstructure parameter c to Neff, indicating that the increase of magnetocrystalline anisotropy field HA and c/Neff can effectively decrease the absolute value of β. On the basis of this analysis, a sintered Nd-Fe-B magnet with a low temperature coefficient of Hcj was fabricated through composition design, and the value of β was only -0.385%/ºC in the temperature interval of 20-150ºC.

  6. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  7. Physics basis and mechanical design of the actively cooled duct scraper protection for the JET neutral beam enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, D.J. [UKAEA Fusion/Euratom Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)], E-mail: dwilson@jet.uk; Ciric, D.; Cox, S.J.; Jones, T.T.C.; Kovari, M. [UKAEA Fusion/Euratom Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [Association EURATOM-CEA, CEA-Cadarache, 13108 St. Paul-Lez-Durance (France); Martin, D.; Milnes, J.; Shannon, M.; Surrey, E. [UKAEA Fusion/Euratom Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2007-10-15

    The objectives of the JET neutral beam enhancement (NBE) include raising the delivered power from the present 25 MW to more than 34 MW and increasing the pulse length from 10 to 20 s. The additional power will be obtained partly by increasing the fractional energy components of the beam, resulting from acceleration of molecular ions, hence increasing the total particle flux. These changes place extreme demands on the design of the upgraded protection to the torus entry duct. The present inertial duct protection already reaches its thermomechanical limit in 10 s pulses, and active cooling of the upgraded duct protection is therefore essential. Extensive analysis of the pressure and temperature evolution in the present un-cooled duct established the relationship between gas re-emission and surface temperature for copper in this operating environment. This information was used in an integrated physics and engineering approach to the design of the actively cooled duct protection, taking into account the power loads from direct beam interception and re-ionisation. Surface temperature determines power density through the gas re-emission and consequential beam re-ionisation. These considerations define the normal operating point for the chosen enhanced hypervapotron element technology. This approach demonstrated that supplementary in situ duct cryopumping would not be needed, provided that the required heat-transfer performance could be met without any encroachment of the elements beyond the space envelope of the existing inertial duct protection plates. This requirement posed severe constraints on the mechanical design of the hypervapotron element array and its manifolding; the adopted engineering design solutions are presented.

  8. Conceptual design considerations for providing hook-up type schemes for tracking beyond design basis events (BDBE) for 700 MWe PHWR project

    International Nuclear Information System (INIS)

    Vhora, S.F.; Inder Jit; Bhardwaj, S.A.

    2005-01-01

    A broad review of major nuclear accidents such as Chernobyl reveals that provision of access to the reactor core for cooling purpose had to be made from outside the reactor building by tunneling. Also the NAPS fire incident could be mitigated once the fire water injection to the steam generators could be ensured. In this case the boiler room which was outside the primary containment was accessible relatively easily for mitigation after the initial period. Both of the above had accident scenarios which can be termed Beyond Design Basis (BDBE) since the accident initiation/scenario did not fit into the events under postulated initiating events (PIES) or Design Basis Events (DBEs). These accidents or events reveal that some sort of access to the core or the components inside the Reactor building becomes necessary. It is also to be noted that manual intervention beyond the initial period of half an hour or earlier in the Emergency operating procedure (EOP) is inevitably called for as a recovery action in order to mitigate the severity and minimize long term consequences. This paper attempts to discuss the type of concepts which can give access to the core or associated systems which can then provide continued heat sink. The discussions would include the criteria for design of such concepts and give examples of such concepts already implemented and proposes schemes to be implemented in the 700 MWe Project. (author)

  9. An Innovative High-Tech Acupuncture Product: SXDZ-100 Nerve Muscle Stimulator, Its Theoretical Basis, Design, and Application

    Directory of Open Access Journals (Sweden)

    Xinyan Gao

    2012-01-01

    Full Text Available We introduce the theoretical basis, design, and application of a patented innovative high-tech product, SXDZ-100 nerve and muscle stimulator. This product is featured with a built-in chip containing transcoding information from different acupuncture manipulation collected from the wide dynamic neurons (WDR in the spinal dorsal horn in animal experiments, which is bioinformation feedback therapy. The discharges of WDR neurons excited by different manipulations are analyzed using chaos theory in this study. It combines the advantages of manual acupuncture (MA like no receptor adaptation and treatment individualization and that of electroacupuncture (EA such as relatively low stimulation intensity and good quantification and thus makes it more effective than common stimulators in acupuncture clinic.

  10. Design basis earthquakes for critical industrial facilities and their characteristics, and the Southern Hyogo prefecture earthquake, 17 January 1995

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Heki

    1998-12-01

    This paper deals with how to establish the concept of the design basis earthquake (DBE) for critical industrial facilities such as nuclear power plants in consideration of disasters such as the Southern Hyogo prefecture earthquake, the so-called Kobe earthquake in 1995. The author once discussed various DBEs at the 7th World Conference on Earthquake Engineering. At that time, the author assumed that the strongest effective PGA would be 0.7 G, and compared the values of accelerations of a structure obtained by various codes in Japan and other countries. The maximum PGA observed by an instrument at the Southern Hyogo prefecture earthquake in 1995 exceeded the previous assumption of the author, even though the results of the previous paper had been pessimistic. According to the experience of the Kobe event, the author will point out the necessity of the third earthquake S{sub s} adding to S{sub 1} and S{sub 2} of previous DBEs.

  11. Design basis threat analysis and implementation of the physical protection system at Nuclear Facility of BATAN Yogyakarta

    International Nuclear Information System (INIS)

    Syarip

    2005-01-01

    An analysis to determine the design basis threat (DBT) and its follow-up through the implementation of physical protection system at the nuclear facility of BATAN Yogyakarta has been done. Methodology used for the analysis is based on the IAEA guidance for the development and maintenance of a DBT. Based on the analysis results, it can be concluded that the threat motivation is influenced by political situation (related to the government policy), criminal, sabotage and theft. The characteristics of threats are: not so well organized, terror, theft of materials information, involving insider (collusion), and intimidation to workers. Potential threat could from guests/students who take a practical job or laboratory exercise. Therefore, it is necessary to be anticipated the possibility and its impact of turmoil/demonstrators such as destruction of: lighting, road, fence, sabotage on the electric and communication lines, surrounding the Yogyakarta nuclear facility

  12. After Action Report:Idaho National Laboratory (INL) 2014 Multiple Facility Beyond Design Basis (BDBE) Evaluated Drill October 21, 2014

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, V. Scott [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-12-01

    On October 21, 2014, Idaho National Laboratory (INL), in coordination with local jurisdictions, and Department of Energy (DOE) Idaho Operations Office (DOE ID) conducted an evaluated drill to demonstrate the ability to implement the requirements of DOE O 151.1C, “Comprehensive Emergency Management System” when responding to a beyond design basis event (BDBE) scenario as outlined in the Office of Health, Safety, and Security Operating Experience Level 1 letter (OE-1: 2013-01). The INL contractor, Battelle Energy Alliance, LLC (BEA), in coordination with CH2M-WG Idaho, LLC (CWI), and Idaho Treatment Group LLC (ITG), successfully demonstrated appropriate response measures to mitigate a BDBE event that would impact multiple facilities across the INL while protecting the health and safety of personnel, the environment, and property. Offsite response organizations participated to demonstrate appropriate response measures.

  13. Calculation of stricken to mortality and incidence cancers due to beyond design basis accidents of the Esfahan Fuel Production Factory

    International Nuclear Information System (INIS)

    Heydari Azar, A.; Shahshahani, M.; Roshanzamir, M.; Sabouhi, R.

    2008-01-01

    In this investigation the amount of absorbed doses by the different pathways of Cloud shine, Ground shine, deposition of radioactive materials on skin and cloths, ingestion, inhalation and the consequences of radioactive material releases due to Beyond Design Basis Accidents such as fire, sintering furnace explosion, criticality and earthquake in Esfahan Fuel Production factory by the residents are evaluated. The calculations related to atomic cloud distribution, estimation of delivered dose and decay chains are performed by PCCOSYMA dose. These computations are based on radioactive source terms, distribution height of radioactive materials. actions for reducing the absorbed dose, human body physiological characteristics, metrological condition and population distribution. Finally, the number of peoples who are stricken to mortality and morbidity cancers and risk values are calculated for 1 year and 50 years

  14. Extending the application range of a fuel performance code from normal operating to design basis accident conditions

    International Nuclear Information System (INIS)

    Van Uffelen, P.; Gyori, C.; Schubert, A.; Laar, J. van de; Hozer, Z.; Spykman, G.

    2008-01-01

    Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events

  15. Calculation of particulate dispersion in a design-basis tornadic storm from the Battelle Memorial Institute, Columbus, Ohio

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1980-10-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Battelle Memorial Institute at Columbus, Ohio. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind value is based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume

  16. Calculation of particulate dispersion in a design-basis tornadic storm from the Babcock and Wilcox Plant, Leechburg, Pennsylvania

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1978-03-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Babcock and Wilcox Plutonium Fabrication Facility at Leechburg, Pennsylvania. Plutonium particles lss than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The quasi-Lagrangian method of moments is used to model the transport of concentration within a grid cell volume

  17. Calculation of particulate dispersion in a design-basis tornadic storm from the Battelle Memorial Institute, Columbus, Ohio

    Energy Technology Data Exchange (ETDEWEB)

    Pepper, D.W.

    1980-10-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Battelle Memorial Institute at Columbus, Ohio. Plutonium particles less than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind value is based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume.

  18. Calculation of particulate dispersion in a design-basis tornadic storm from the Babcock and Wilcox Plant, Leechburg, Pennsylvania

    Energy Technology Data Exchange (ETDEWEB)

    Pepper, D.W.

    1978-03-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Babcock and Wilcox Plutonium Fabrication Facility at Leechburg, Pennsylvania. Plutonium particles lss than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The quasi-Lagrangian method of moments is used to model the transport of concentration within a grid cell volume.

  19. Subnanodimensional thermometrical NMR-sensors on the basis of lanthanide(III) paramagnetic complexes with EDTA for temperature control in aqueous media and magnetoresonance tomography

    International Nuclear Information System (INIS)

    Babajlov, S.P.

    2008-01-01

    It is proposed that temperature dependence of paramagnetic lanthanide-induced shifts (LIS) in NMR spectra on nuclei of EDTA type synthetic organic complexes in kinetically unstable compounds with paramagnetic lanthanide(III) cations is used for ascertaining the temperature of samples placed directly into a NMR spectrometer and formed on the basis of aqueous solutions of diverse chemical substances. It was revealed that complex [Ho III (EDTA)] can be used as an internal or an external thermometric NMR-sensor. For identification and control of temperature in a sample one can make use of LIS for individual signals from CH 2 groups (taken in relation to water or inner DCC standard signals). A higher temperature measurement accuracy (≤0.08 K) is attained by using LIS difference corresponding to the relevant nonequivalent CH 2 groups [ru

  20. Analysis of the metallic containment integrity of Angra 2/3 reactor under the effects of the design basis accident

    International Nuclear Information System (INIS)

    Costa, J.R.

    1981-06-01

    The application of Condru 4 computer code, developed to determine the maximum values of pressure and temperature that occur inside the metallic containment building of PWR nuclear power plants, in case of a hypothetic accident - LOCA - considered as a Design Basic Accident - DBA. The hypothesis, input and results for the simulation of a loss of coolant in the hot leg of the Angra-2/3 reactors, considered as the most critical case for that Kind of project, are presented. The analysis was made with input provided by the manufacturer. (Author) [pt

  1. Practical design constraints for using secondary concentrators at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    O' Gallagher, J.J.; Winston, R.

    1999-07-01

    The optical advantages of using nonimaging secondary concentrators in two-stage solar thermal dish systems are well understood. However, practical questions having to do with the thermal behavior of any secondary and its possible effects on the performance of cavity type receivers have only recently begun to be investigated. A few years ago an experimental demonstration of a trumpet type nonimaging secondary concentrator was carried out with a cavity receiver operating 660 C in combination with the Cummins Power Generation CPG-460 7.5 kWe concentrator system. Lessons learned from this and previous experiments are reviewed. The tests alleviated any operational concerns about the effectiveness of active water cooling and have shown that secondaries can be operated successfully at high temperatures without significant problems. There was no evidence of direct heat loss from the hot receiver to the cooled trumpet. The optical quality of any primary can be expected to fall well below design goals and to deteriorate further with time. This expectation should be taken into account in planning future experiments and developing new concentrating systems.

  2. Control room unfiltered in-leakage limit analysis of design-basis LOCA for Lungmen ABWR plant

    International Nuclear Information System (INIS)

    Tsai Chihming; Chang Chinjang; Yuann Yngruey

    2014-01-01

    In USNRC's Generic Letter 2003-01, 'Control Room Habitability,' it requests utilities provide information to demonstrate that the control room at each of their respective facilities complies with the current licensing and design bases, and applicable regulatory requirements, and that suitable design, maintenance and testing control measures are in place for maintaining this compliance. In particular, each utility is required to perform the control room in-leakage test to demonstrate that the unfiltered in-leakage rate is within that assumed in the licensing analyses. It must be ensured that the control room envelope habitability, in terms of radiation dose, is maintained during normal operations as well as design basis accidents. In view of this, a dose analysis has been performed to establish the control room unfiltered in-leakage limit which can be used as an acceptance criterion for the in-leakage test. The analysis in this study is for Lungmen ABWR plant. The plant has twin units, with each unit having its own control room. The TID-4844 source terms and associated methodology are used. The USNRC RADTRAD v3.03 code is employed for the transport calculation of radioactive materials in different paths, including control room in-leakage path. The radiological criterion on protection of the operators specified in 10 CFR 50, Appendix A, General Design Criterion 19 is followed. It's demonstrated that the performance of Lungmen control room with 500 cfm unfiltered in-leakage air could meet the radiological habitability acceptance criteria in case of radiation hazards. (author)

  3. Large-Scale Atmospheric Circulation Patterns Associated with Temperature Extremes as a Basis for Model Evaluation: Methodological Overview and Results

    Science.gov (United States)

    Loikith, P. C.; Broccoli, A. J.; Waliser, D. E.; Lintner, B. R.; Neelin, J. D.

    2015-12-01

    Anomalous large-scale circulation patterns often play a key role in the occurrence of temperature extremes. For example, large-scale circulation can drive horizontal temperature advection or influence local processes that lead to extreme temperatures, such as by inhibiting moderating sea breezes, promoting downslope adiabatic warming, and affecting the development of cloud cover. Additionally, large-scale circulation can influence the shape of temperature distribution tails, with important implications for the magnitude of future changes in extremes. As a result of the prominent role these patterns play in the occurrence and character of extremes, the way in which temperature extremes change in the future will be highly influenced by if and how these patterns change. It is therefore critical to identify and understand the key patterns associated with extremes at local to regional scales in the current climate and to use this foundation as a target for climate model validation. This presentation provides an overview of recent and ongoing work aimed at developing and applying novel approaches to identifying and describing the large-scale circulation patterns associated with temperature extremes in observations and using this foundation to evaluate state-of-the-art global and regional climate models. Emphasis is given to anomalies in sea level pressure and 500 hPa geopotential height over North America using several methods to identify circulation patterns, including self-organizing maps and composite analysis. Overall, evaluation results suggest that models are able to reproduce observed patterns associated with temperature extremes with reasonable fidelity in many cases. Model skill is often highest when and where synoptic-scale processes are the dominant mechanisms for extremes, and lower where sub-grid scale processes (such as those related to topography) are important. Where model skill in reproducing these patterns is high, it can be inferred that extremes are

  4. Assessment of precast beam-column using capacity demand response spectrum subject to design basis earthquake and maximum considered earthquake

    Science.gov (United States)

    Ghani, Kay Dora Abd.; Tukiar, Mohd Azuan; Hamid, Nor Hayati Abdul

    2017-08-01

    Malaysia is surrounded by the tectonic feature of the Sumatera area which consists of two seismically active inter-plate boundaries, namely the Indo-Australian and the Eurasian Plates on the west and the Philippine Plates on the east. Hence, Malaysia experiences tremors from far distant earthquake occurring in Banda Aceh, Nias Island, Padang and other parts of Sumatera Indonesia. In order to predict the safety of precast buildings in Malaysia under near field ground motion the response spectrum analysis could be used for dealing with future earthquake whose specific nature is unknown. This paper aimed to develop of capacity demand response spectrum subject to Design Basis Earthquake (DBE) and Maximum Considered Earthquake (MCE) in order to assess the performance of precast beam column joint. From the capacity-demand response spectrum analysis, it can be concluded that the precast beam-column joints would not survive when subjected to earthquake excitation with surface-wave magnitude, Mw, of more than 5.5 Scale Richter (Type 1 spectra). This means that the beam-column joint which was designed using the current code of practice (BS8110) would be severely damaged when subjected to high earthquake excitation. The capacity-demand response spectrum analysis also shows that the precast beam-column joints in the prototype studied would be severely damaged when subjected to Maximum Considered Earthquake (MCE) with PGA=0.22g having a surface-wave magnitude of more than 5.5 Scale Richter, or Type 1 spectra.

  5. Design of cognitive engine for cognitive radio based on the rough sets and radial basis function neural network

    Science.gov (United States)

    Yang, Yanchao; Jiang, Hong; Liu, Congbin; Lan, Zhongli

    2013-03-01

    Cognitive radio (CR) is an intelligent wireless communication system which can dynamically adjust the parameters to improve system performance depending on the environmental change and quality of service. The core technology for CR is the design of cognitive engine, which introduces reasoning and learning methods in the field of artificial intelligence, to achieve the perception, adaptation and learning capability. Considering the dynamical wireless environment and demands, this paper proposes a design of cognitive engine based on the rough sets (RS) and radial basis function neural network (RBF_NN). The method uses experienced knowledge and environment information processed by RS module to train the RBF_NN, and then the learning model is used to reconfigure communication parameters to allocate resources rationally and improve system performance. After training learning model, the performance is evaluated according to two benchmark functions. The simulation results demonstrate the effectiveness of the model and the proposed cognitive engine can effectively achieve the goal of learning and reconfiguration in cognitive radio.

  6. Pressure vessel design codes: A review of their applicability to HTGR components at temperatures above 800 deg C

    International Nuclear Information System (INIS)

    Hughes, P.T.; Over, H.H.; Bieniussa, K.

    1984-01-01

    The governments of USA and Federal Republic of Germany have approved of cooperation between the two countries in an endeavour to establish structural design code for gas reactor components intended to operate at temperatures exceeding 800 deg C. The basis of existing codes and their applicability to gas reactor component design are reviewed in this paper. This review has raised a number of important questions as to the direct applicability of the present codes. The status of US and FRG cooperative efforts to obtain answers to these questions are presented

  7. Development of Emergency Operating Strategies for Beyond Design Basis External Event(BDBEE)s in Korean WH Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Duk-Joo; Lee, Seung-Chan; Sung, Je-Joong; Ha, Sang-Jun [KHNP CRI, Daejeon (Korea, Republic of); Hong, Soon-Joon; Hwang, Su-Hyun; Lee, Byung-Chul; Park, Kang-Min [FNC Tech. Co., Yongin (Korea, Republic of)

    2015-10-15

    Westinghouse developed and connected emergency operating procedures into a set of FLEX Support Guidelines(FSGs). This paper explains that Korean WH(Westinghouse) type nuclear power plants develop emergency operating strategies for ELAP(extended loss of all AC power), which include guidelines to use permanent and portable equipment as necessary to prevent core damage until AC power is restored from a reliable alternate source of AC power. The Korean emergency operating response strategies were developed to cope with a ELAP such as Fukushima event. The strategies include guidelines to prevent fuel damage using the FLEX equipment. Operators should take actions to prepare FLEX equipment within license basis SBO coping time. The loss of all AC power has been analyzed to identify the behavior of major NSSS process variables using RELAP computer code. The accident analysis showed that the plant does not result in fuel damage in 72 hours after an ELAP if operators take actions to cool RCS with opening of SG ADV in 5 gpm seal leak case. In this scenario, because ELAP is in process and all power cannot be used, operator should operate the FLEX equipment in order to actuate active equipment using the EOP fo SBO response. This strategy will prevent entering SAMG because this actions result in core cooling and stay in core exit temperature less than 650 .deg. C. Korean emergency operating guidelines(EOGs) will be developed using this strategies for response to the BDBEE.

  8. Design of indoor temperature and humidity detection system based on single chip microcomputer

    Science.gov (United States)

    Fu, Xiuwei; Fu, Li; Ma, Tianhui

    2018-03-01

    The indoor temperature and humidity detection system based on STC15F2K60S2 is designed in this paper. The temperature and humidity sensor DHT22 to monitor the indoor temperature and humidity are used, and the temperature and humidity data to the user's handheld device are wirelessly transmitted, when the temperature reaches or exceeds the user set the temperature alarm value, and the system sound and light alarm, to remind the user.

  9. Design considerations for CRBRP heat transport system piping operating at elevated temperatures

    International Nuclear Information System (INIS)

    Pollono, L.P.; Mello, R.M.

    1979-01-01

    The heat transport system sodium piping for the Clinch River Breeder Reactor Plant (CRBRP) within the reactor containment building must withstand high temperatures for long periods of time. Each phase of the mechanical design process of the piping system is influenced by elevated temperature considerations which include material thermal creep effects, ratchetting caused by rapid temperature transients and stress relaxation, and material degradation effects. The structural design philosophy taken to design the CRBRP piping operating in a high temperature environment is described. The resulting design of the heat transport system piping is presented along with a discussion of special features that resulted from the elevated temperature considerations

  10. Design of shell-and-tube heat exchangers when the fouling depends on local temperature and velocity

    Energy Technology Data Exchange (ETDEWEB)

    Butterworth, D. [HTFS, Hyprotech, Didcot (United Kingdom)

    2002-07-01

    Shell-and-tube heat exchangers are normally designed on the basis of a uniform and constant fouling resistance that is specified in advance by the exchanger user. The design process is then one of determining the best exchanger that will achieve the thermal duty within the specified pressure drop constraints. It has been shown in previous papers [Designing shell-and-tube heat exchangers with velocity-dependant fouling, 34th US national Heat Transfer Conference, 20-22 August 2000, Pittsburg, PA; Designing shell-and-tube heat exchangers with velocity-dependant fouling, 2nd Int. Conf. on Petroleum and Gas Phase Behavior and Fouling, 27-31 August 2000, Copenhagen] that this approach can be extended to the design of exchangers where the design fouling resistance depends on velocity. The current paper briefly reviews the main findings of the previous papers and goes on to treat the case where the fouling depends also on the local temperatures. The Ebert-Panchal [Analysis of Exxon crude-oil, slip-stream coking data, Engineering Foundation Conference on Fouling Mitigation of Heat Exchangers, 18-23 June 1995, California] form of fouling rate equation is used to evaluate this fouling dependence. When allowing for temperature effects, it becomes difficult to divorce the design from the way the exchanger will be operated up to the point when the design fouling is achieved. However, rational ways of separating the design from the operation are proposed. (author)

  11. A transcriptomic analysis of bermudagrass (Cynodon dactylon) provides novel insights into the basis of low temperature tolerance.

    Science.gov (United States)

    Chen, Liang; Fan, Jibiao; Hu, Longxing; Hu, Zhengrong; Xie, Yan; Zhang, Yingzi; Lou, Yanhong; Nevo, Eviatar; Fu, Jinmin

    2015-09-11

    Cold stress is regarded as a key factor limiting widespread use for bermudagrass (Cynodon dactylon). Therefore, to improve cold tolerance for bermudagrass, it is urgent to understand molecular mechanisms of bermudagrass response to cold stress. However, our knowledge about the molecular responses of this species to cold stress is largely unknown. The objective of this study was to characterize the transcriptomic response to low temperature in bermudagrass by using RNA-Seq platform. Ten cDNA libraries were generated from RNA samples of leaves from five different treatments in the cold-resistant (R) and the cold-sensitive (S) genotypes, including 4 °C cold acclimation (CA) for 24 h and 48 h, freezing (-5 °C) treatments for 4 h with or without prior CA, and controls. When subjected to cold acclimation, global gene expressions were initiated more quickly in the R genotype than those in the S genotype. The R genotype activated gene expression more effectively in response to freezing temperature after 48 h CA than the S genotype. The differentially expressed genes were identified as low temperature sensing and signaling-related genes, functional proteins and transcription factors, many of which were specifically or predominantly expressed in the R genotype under cold treatments, implying that these genes play important roles in the enhanced cold hardiness of bermudagrass. KEGG pathway enrichment analysis for DEGs revealed that photosynthesis, nitrogen metabolism and carbon fixation pathways play key roles in bermudagrass response to cold stress. The results of this study may contribute to our understanding the molecular mechanism underlying the responses of bermudagrass to cold stress, and also provide important clues for further study and in-depth characterization of cold-resistance breeding candidate genes in bermudagrass.

  12. Japan Catastrophic Earthquake and Tsunami in Fukushima Daiichi NPP; Is it Beyond Design Basis Accident or a Design Deficiency and Operator Unawareness?

    International Nuclear Information System (INIS)

    Gaafar, M.A.; Refeat, R.M.; EL-Kady, A.A.

    2012-01-01

    On March 11, 2011 a catastrophic earthquake and tsunami struck the north east coast of Japan. This catastrophe damaged fully or partially the six units of the Fukushima Daiichi Nuclear Power Plant.Questions were raised following the aftermath, whether it is beyond design basis accident caused by severe natural event or a failure by the Japanese authorities to plan to deal with such accident. There are many indications that the Utility of Fukushima Daiichi NPP, Tokyo Electric Power Company (TEPCO), did not pay enough attention to numerous facts about the incompatibility of the site and several design defects in the plant units. In fact there are three other NPP sites nearby Fukushima Daiichi Plant (about 30 to 60 Km far from Fukushima Daiichi NPP), with different site characteristics, which survived the same catastrophic earthquake and tsunami, but they were automatically turned into a safe shutdown state. These plants sites are Fukushima Daini Plant (4 units), Onagawa Plant (3 units) and Tokai Daini (II) Plant (one unit). In this paper, the aftermath Fukushima Daiichi plant integrity is pointed out. Some facts about the site and design concerns which could have implications on the accident are discussed. The response of Japan Authority is outlined and some remarks about their actions are underlined. The impacts of this disaster on the Nuclear Power Program worldwide are also discussed.

  13. Calculation of particulate dispersion in a design-basis tornadic storm from the General Electric Vallecitos Nuclear Center, Vallecitos, California

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1979-11-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the General Electric Vallecitos Nuclear Center at Vallecitos, California. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

  14. Calculation of particulate dispersion in a design-basis tornadic storm from the Exxon Nuclear Company, Richland, Washington

    Energy Technology Data Exchange (ETDEWEB)

    Pepper, D.W.

    1978-07-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Exxon Nuclear Company at Richland, Washington. Plutonium particles less than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm.

  15. Calculation of particulate dispersion in a design-basis tornadic storm from the Exxon Nuclear Company, Richland, Washington

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1978-07-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Exxon Nuclear Company at Richland, Washington. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

  16. Calculation of particulate dispersion in a design-basis tornadic storm from the General Electric Vallecitos Nuclear Center, Vallecitos, California

    Energy Technology Data Exchange (ETDEWEB)

    Pepper, D.W.

    1979-11-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the General Electric Vallecitos Nuclear Center at Vallecitos, California. Plutonium particles less than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm.

  17. Recent UK research and the development of high temperature design methods

    International Nuclear Information System (INIS)

    Rose, R.T.; Tomkins, B.; Townley, C.H.A.

    1987-01-01

    The paper outlines recent research and development activities on high temperature design methods and criteria for high temperature components as utilized by liquid metal cooled fast breeder reactors. (orig.)

  18. Electronic Modeling and Design for Extreme Temperatures, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to develop electronics for operation at temperatures that range from -230oC to +130oC. This new technology will minimize the requirements for external...

  19. The Design of Temperature and Humidity Chamber Monitor and Controller

    OpenAIRE

    Tibebu, Simachew

    2016-01-01

    The temperature and humidity chamber, (climate chamber) is a device located at the Technobothnia Education and Research Center that simulates different climate conditions. The simulated environment is used to test the capabilities of electrical equipment in different temperature and humidity conditions. The climate chamber, among other things houses a dedicated computer, the control PC, and a control software running in it which together are responsible for running and control-ling these simu...

  20. Opto-mechanical subsystem with temperature compensation through isothemal design

    Science.gov (United States)

    Goodwin, F. E. (Inventor)

    1977-01-01

    An opto-mechanical subsystem for supporting a laser structure which minimizes changes in the alignment of the laser optics in response to temperature variations is described. Both optical and mechanical structural components of the system are formed of the same material, preferably beryllium, which is selected for high mechanical strength and good thermal conducting qualities. All mechanical and optical components are mounted and assembled to provide thorough thermal coupling throughout the subsystem to prevent the development of temperature gradients.

  1. Design of a low-cost system for electrical conductivity measurements of high temperature

    Science.gov (United States)

    Singh, Yadunath

    2018-05-01

    It is always a curiosity and interest among researchers working in the field of material science to know the impact of high temperature on the physical and transport properties of the materials. In this paper, we report on the design and working of a system for the measurements of electrical resistivity with high temperature. It was designed at our place and successively used for these measurements in the temperature range from room temperature to 500 ˚C.

  2. Design, Qualification and Integration Testing of the High-Temperature Resistance Temperature Device for Stirling Power System

    Science.gov (United States)

    Chan, Jack; Hill, Dennis H.; Elisii, Remo; White, Jonathan R.; Lewandowski, Edward J.; Oriti, Salvatore M.

    2015-01-01

    The Advanced Stirling Radioisotope Generator (ASRG), developed from 2006 to 2013 under the joint sponsorship of the United States Department of Energy (DOE) and National Aeronautics and Space Administration (NASA) to provide a high-efficiency power system for future deep space missions, employed Sunpower Incorporated's Advanced Stirling Convertors (ASCs) with operating temperature up to 840 C. High-temperature operation was made possible by advanced heater head materials developed to increase reliability and thermal-to-mechanical conversion efficiency. During a mission, it is desirable to monitor the Stirling hot-end temperature as a measure of convertor health status and assist in making appropriate operating parameter adjustments to maintain the desired hot-end temperature as the radioisotope fuel decays. To facilitate these operations, a Resistance Temperature Device (RTD) that is capable of high-temperature, continuous long-life service was designed, developed and qualified for use in the ASRG. A thermal bridge was also implemented to reduce the RTD temperature exposure while still allowing an accurate projection of the ASC hot-end temperature. NASA integrated two flight-design RTDs on the ASCs and assembled into the high-fidelity Engineering Unit, the ASRG EU2, at Glenn Research Center (GRC) for extended operation and system characterization. This paper presents the design implementation and qualification of the RTD, and its performance characteristics and calibration in the ASRG EU2 testing.

  3. TECHNICAL BASIS FOR DOE STANDARD 3013 EQUIVALENCY SUPPORTING REDUCED TEMPERATURE STABILIZATION OF OXALATE-DERIVED PLUTONIUM OXIDE PRODUCED BY THE HB-LINE FACILITY AT SAVANNAH RIVER SITE

    Energy Technology Data Exchange (ETDEWEB)

    Duffey, J.; Livingston, R.; Berg, J.; Veirs, D.

    2012-07-02

    The HB-Line (HBL) facility at the Savannah River Site (SRS) is designed to produce high-purity plutonium dioxide (PuO{sub 2}) which is suitable for future use in production of Mixed Oxide (MOX) fuel. The MOX Fuel Fabrication Facility (MFFF) requires PuO{sub 2} feed to be packaged per the U.S. Department of Energy (DOE) Standard 3013 (DOE-STD-3013) to comply with the facility's safety basis. The stabilization conditions imposed by DOE-STD-3013 for PuO{sub 2} (i.e., 950 C for 2 hours) preclude use of the HBL PuO{sub 2} in direct fuel fabrication and reduce the value of the HBL product as MFFF feedstock. Consequently, HBL initiated a technical evaluation to define acceptable operating conditions for production of high-purity PuO{sub 2} that fulfills the DOE-STD-3013 criteria for safe storage. The purpose of this document is to demonstrate that within the defined operating conditions, the HBL process will be equivalent for meeting the requirements of the DOE-STD-3013 stabilization process for plutonium-bearing materials from the DOE complex. The proposed 3013 equivalency reduces the prescribed stabilization temperature for high-purity PuO{sub 2} from oxalate precipitation processes from 950 C to 640 C and places a limit of 60% on the relative humidity (RH) at the lowest material temperature. The equivalency is limited to material produced using the HBL established flow sheet, for example, nitric acid anion exchange and Pu(IV) direct strike oxalate precipitation with stabilization at a minimum temperature of 640 C for four hours (h). The product purity must meet the MFFF acceptance criteria of 23,600 {micro}g/g Pu (i.e., 2.1 wt %) total impurities and chloride content less than 250 {micro}g/g of Pu. All other stabilization and packaging criteria identified by DOE-STD-3013-2012 or earlier revisions of the standard apply. Based on the evaluation of test data discussed in this document, the expert judgment of the authors supports packaging the HBL product under a 3013

  4. Evaluation on Cooling Performance of Containment Fan Cooler during Design Basis Accident with Loss of Offsite Power for Kori 3 and 4 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Bok; Lee, Sang Won [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Park, Young Chan [Atomic Creative Technology Co., LTD., Daejeon (Korea, Republic of)

    2007-10-15

    The purpose of this study is to evaluate cooling performance of containment fan cooler units and to review a technical background related to Generic Letter 96-06. In case that design basis accident (DBA) and loss of offsite power (LOOP) occurs, component cooling water (CCW) pumps cannot provide the cooling water source to fan cooler units while fan coolers coast down. Fan cooler units and CCW pumps are restarted by emergency diesel generator (EDG) operation and it takes about 30 seconds. In this scenario, before the EDG restarts and CCW flowrate is restored, heated air in the containment passes through coil of fan cooler units without cooling water source. In this situation, the boiling of water in the fan cooler units may occur. Restarting of CCW pumps may bring about condensation by injected cooling water and water hammer may occur. This thermal-hydraulic effect is sensitive to system configuration, i.e system pressure, containment pressure/temperature, EDG restarting time, etc. In this study, the evaluation of containment fan cooler units was performed for Kori 3 and 4 nuclear power plant.

  5. design and analysis of a multipoint temperature datalogger

    African Journals Online (AJOL)

    user

    The system gives room for effective and adjustable temperature data logging procedure. Hence ... 120.00. 11. Soldering Board. 3. 360.00. 360.00. 12. Rainbow flex wires. 1m. 500.00 ... An alcohol thermometer was used as a control. In order.

  6. Design of a CMOS temperature sensor with current output

    NARCIS (Netherlands)

    Kolling, A.; Kölling, Arjan; Bak, Frans; Bergveld, Piet; Seevinck, E.; Seevinck, Evert

    1990-01-01

    In this paper a CMOS temperature-to-current converter is presented of which the output current is the difference between a PTC current and an NTC current. The PTC current is derived from a PTAT cell, while the NTC current is derived from a threshold voltage reference source. It is shown that this

  7. High-temperature-structural design and research and development for reactor system components

    International Nuclear Information System (INIS)

    Matsumura, Makoto; Hada, Mikio

    1985-01-01

    The design of reactor system components requires high-temperature-structural design guide with the consideration of the creep effect of materials related to research and development on structural design. The high-temperature-structural design guideline for the fast prototype reactor MONJU has been developed under the active leadership by Power Reactor and Nuclear Fuel Development Corporation and Toshiba has actively participated to this work with responsibility on in-vessel components, performing research and development programs. This paper reports the current status of high-temperature-structural-design-oriented research and development programs and development of analytical system including stress-evaluation program. (author)

  8. Guidelines for the structural design of experimental multi-purpose VHTR at the elevated temperature services

    International Nuclear Information System (INIS)

    Nomura, Sueo; Uga, Takeo; Miyamoto, Yoshiaki; Muto, Yasushi; Ikushima, Takeshi

    1976-02-01

    The guidelines are presented for structural design of the experimental multi-purpose VHTR(Very High Temperature Reactor) at the elevated temperature services. Covered are features of the VHTR structural design, specifications, safety design, seismic design, failure modes to be considered, stress criteria for various load combinations and the mechanical properties of the materials. The guidelines were prepared by referring to safety criteria of high-temperature gas cooled reactors, ASME Boiler and Pressure Vessel code, Section III, case 1592 and the domestic seismic design guide of nuclear power facilities. (auth.)

  9. HYFIRE II: fusion/high-temperature electrolysis conceptual-design study. Annual report

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1983-08-01

    As in the previous HYFIRE design study, the current study focuses on coupling a Tokamak fusion reactor with a high-temperature blanket to a High-Temperature Electrolyzer (HTE) process to produce hydrogen and oxygen. Scaling of the STARFIRE reactor to allow a blanket power to 6000 MW(th) is also assumed. The primary difference between the two studies is the maximum inlet steam temperature to the electrolyzer. This temperature is decreased from approx. 1300 0 to approx. 1150 0 C, which is closer to the maximum projected temperature of the Westinghouse fuel cell design. The process flow conditions change but the basic design philosophy and approaches to process design remain the same as before. Westinghouse assisted in the study in the areas of systems design integration, plasma engineering, balance-of-plant design, and electrolyzer technology

  10. Material Performance of Fully-Ceramic Micro-Encapsulated Fuel under Selected LWR Design Basis Scenarios: Final Report

    International Nuclear Information System (INIS)

    Boer, B.; Sen, R.S.; Pope, M.A.; Ougouag, A.M.

    2011-01-01

    The extension to LWRs of the use of Deep-Burn coated particle fuel envisaged for HTRs has been investigated. TRISO coated fuel particles are used in Fully-Ceramic Microencapsulated (FCM) fuel within a SiC matrix rather than the graphite of HTRs. TRISO particles are well characterized for uranium-fueled HTRs. However, operating conditions of LWRs are different from those of HTRs (temperature, neutron energy spectrum, fast fluence levels, power density). Furthermore, the time scales of transient core behavior during accidents are usually much shorter and thus more severe in LWRs. The PASTA code was updated for analysis of stresses in coated particle FCM fuel. The code extensions enable the automatic use of neutronic data (burnup, fast fluence as a function of irradiation time) obtained using the DRAGON neutronics code. An input option for automatic evaluation of temperature rise during anticipated transients was also added. A new thermal model for FCM was incorporated into the code; so-were updated correlations (for pyrocarbon coating layers) suitable to estimating dimensional changes at the high fluence levels attained in LWR DB fuel. Analyses of the FCM fuel using the updated PASTA code under nominal and accident conditions show: (1) Stress levels in SiC-coatings are low for low fission gas release (FGR) fractions of several percent, as based on data of fission gas diffusion in UO 2 kernels. However, the high burnup level of LWR-DB fuel implies that the FGR fraction is more likely to be in the range of 50-100%, similar to Inert Matrix Fuels (IMFs). For this range the predicted stresses and failure fractions of the SiC coating are high for the reference particle design (500 (micro)mm kernel diameter, 100 (micro)mm buffer, 35 (micro)mm IPyC, 35 (micro)mm SiC, 40 (micro)mm OPyC). A conservative case, assuming 100% FGR, 900K fuel temperature and 705 MWd/kg (77% FIMA) fuel burnup, results in a 8.0 x 10 -2 failure probability. For a 'best-estimate' FGR fraction of 50

  11. Design study of SMES system using high temperature superconductors

    International Nuclear Information System (INIS)

    Yoshihara, T.; Masuda, M.; Shintomi, T.; Hasegawa, J.

    1988-01-01

    Various studies of high Tc superconductors are being energetically pursued all over the world, since IBM Zurich Research Laboratory reported on the superconducting oxide. A new design using a high Tc superconductor is under study for 5000 MWh, on the assumption that it is available like conventional superconductors. Problems related to the Tc SMES system, mainly thermal insulation, refrigeration system, stability of superconductors, etc., are considered. Some design examples of high Tc SMES system are proposed

  12. Robust design of microelectronics assemblies against mechanical shock, temperature and moisture effects of temperature, moisture and mechanical driving forces

    CERN Document Server

    Wong, E-H

    2015-01-01

    Robust Design of Microelectronics Assemblies Against Mechanical Shock, Temperature and Moisture discusses how the reliability of packaging components is a prime concern to electronics manufacturers. The text presents a thorough review of this important field of research, providing users with a practical guide that discusses theoretical aspects, experimental results, and modeling techniques. The authors use their extensive experience to produce detailed chapters covering temperature, moisture, and mechanical shock induced failure, adhesive interconnects, and viscoelasticity. Useful progr

  13. Design and evaluation of an inexpensive radiation shield for monitoring surface air temperatures

    Science.gov (United States)

    Zachary A. Holden; Anna E. Klene; Robert F. Keefe; Gretchen G. Moisen

    2013-01-01

    Inexpensive temperature sensors are widely used in agricultural and forestry research. This paper describes a low-cost (~3 USD) radiation shield (radshield) designed for monitoring surface air temperatures in harsh outdoor environments. We compared the performance of the radshield paired with low-cost temperature sensors at three sites in western Montana to several...

  14. Elevated temperature weldment behavior as related to nuclear design criteria

    International Nuclear Information System (INIS)

    King, R.T.; Canonico, D.A.; Brinkman, C.R.

    1975-01-01

    Factors affecting the physical and mechanical properties of weldments are reviewed. Data are presented that show wide variability in the properties of weld metal and heat-affected zones both within a given weldment and from weldment to weldment. The weld metal or heat-affected zone may be stronger or weaker, and more or less ductile than the base metal joined. This implies that current design rules, which are presumed to have well-known safety factors based on the properties of base metals, may have other actual safety factors in welded construction, due to the differences between weld metal and base metal properties. Suggestions for obtaining further data that can subsequently be used to improve confidence in design rules or to change design rules for nuclear application are presented

  15. Design of Water Temperature Control System Based on Single Chip Microcomputer

    Science.gov (United States)

    Tan, Hanhong; Yan, Qiyan

    2017-12-01

    In this paper, we mainly introduce a multi-function water temperature controller designed with 51 single-chip microcomputer. This controller has automatic and manual water, set the water temperature, real-time display of water and temperature and alarm function, and has a simple structure, high reliability, low cost. The current water temperature controller on the market basically use bimetal temperature control, temperature control accuracy is low, poor reliability, a single function. With the development of microelectronics technology, monolithic microprocessor function is increasing, the price is low, in all aspects of widely used. In the water temperature controller in the application of single-chip, with a simple design, high reliability, easy to expand the advantages of the function. Is based on the appeal background, so this paper focuses on the temperature controller in the intelligent control of the discussion.

  16. Applications of American design codes for elevated temperature environment

    International Nuclear Information System (INIS)

    Severud, L.K.

    1980-03-01

    A brief summary of the ASME Code rules of Case N-47 is presented. An overview of the typical procedure used to demonstrate Code compliance is provided. Application experience and some examples of detailed inelastic analysis and simplified-approximate methods are given. Recent developments and future trends in design criteria and ASME Code rules are also presented

  17. Off-design performance of a chemical looping combustion (CLC) combined cycle: effects of ambient temperature

    Science.gov (United States)

    Chi, Jinling; Wang, Bo; Zhang, Shijie; Xiao, Yunhan

    2010-02-01

    The present work investigates the influence of ambient temperature on the steady-state off-design thermodynamic performance of a chemical looping combustion (CLC) combined cycle. A sensitivity analysis of the CLC reactor system was conducted, which shows that the parameters that influence the temperatures of the CLC reactors most are the flow rate and temperature of air entering the air reactor. For the ambient temperature variation, three off-design control strategies have been assumed and compared: 1) without any Inlet Guide Vane (IGV) control, 2) IGV control to maintain air reactor temperature and 3) IGV control to maintain constant fuel reactor temperature, aside from fuel flow rate adjusting. Results indicate that, compared with the conventional combined cycle, due to the requirement of pressure balance at outlet of the two CLC reactors, CLC combined cycle shows completely different off-design thermodynamic characteristics regardless of the control strategy adopted. For the first control strategy, temperatures of the two CLC reactors both rise obviously as ambient temperature increases. IGV control adopted by the second and the third strategy has the effect to maintain one of the two reactors' temperatures at design condition when ambient temperature is above design point. Compare with the second strategy, the third would induce more severe decrease of efficiency and output power of the CLC combined cycle.

  18. Fundamental principles for a nuclear design and structural analysis code for HTR components operating at temperatures above 8000C

    International Nuclear Information System (INIS)

    Nickel, H.; Schubert, F.

    1985-01-01

    With reference to the special characteristics of an HTR plant for the supply of nuclear process heat, the investigation of the fundamental principles to form the basis for a high temperature nuclear structural design code has been described. As examples, preliminary design values are proposed for the creep rupture and fatigue behaviour. The linear damage accumulation rule is for practical reasons proposed for the determination of service life, and the difficulties in using this rule are discussed. Finally, using the data obtained in structural analysis, the main areas of investigation which will lead to improvements in the utilization of the materials are discussed. Based on the current information, the working group ''Design Code'' believes that a service life of 70000 h for the heat-exchanging components operating at above 800 0 C can be. (orig.)

  19. Shrinkage Effects of the Conduction Zone in the Electrical Properties of Metal Oxide Nanocrystals: The Basis for Room Temperature Conductometric Gas Sensor

    Directory of Open Access Journals (Sweden)

    M. Manzanares

    2009-01-01

    Full Text Available The influence of charge localized at the surface of minute metal oxide nanocrystals was studied in WO3 and In2O3 nanostructures, which were obtained replicating mesoporous silica templates. Here, it is shown that the very high resistive states observed at room temperature and dark conditions were originated by the total shrinkage of the conductive zone in the inner part of these nanocrystals. On the contrary, at room temperature and under UV illumination, both photogenerated electron-hole pairs and empty surface states generated by photons diminished the negative charge accumulated at the surface, enlarging the conductive zone and, as a consequence, leading to a reduction of the electrical resistance. Under these conditions, empty surface states produced by UV light reacted with oxidizing gaseous molecules. The charge exchange associated to these reactions also affected the size of the inner conductive zone, and leaded to a new steady-state resistance. These chemical, physical and geometrical effects can be used for gas detection, and constitutes the basis for developing novel room temperature conductometric gas sensors responsive to oxidizing species.

  20. High temperature hall effect measurement system design, measurement and analysis

    Science.gov (United States)

    Berkun, Isil

    A reliable knowledge of the transport properties of semiconductor materials is essential for the development and understanding of a number of electronic devices. In this thesis, the work on developing a Hall Effect measurement system with software based data acqui- sition and control for a temperature range of 300K-700K will be described. A system was developed for high temperature measurements of materials including single crystal diamond, poly-crystalline diamond, and thermoelectric compounds. An added capability for monitor- ing the current versus voltage behavior of the contacts was used for studying the influence of ohmic and non-ohmic contacts on Hall Effect measurements. The system has been primar- ily used for testing the transport properties of boron-doped single crystal diamond (SCD) deposited in a microwave plasma-assisted chemical vapor deposition (MPCVD) reactor [1]. Diamond has several outstanding properties that are of high interest for its development as an electronic material. These include a relatively wide band gap of 5.5 (eV), high thermal conductivity, high mobility, high saturation velocity, and a high breakdown voltage. For a temperature range of 300K-700K, IV curves, Hall mobilities and carrier concentrations are shown. Temperature dependent Hall effect measurements have shown carrier concentrations from below 1017cm --3 to approximately 1021 cm--3 with mobilities ranging from 763( cm2/V s) to 0.15(cm 2/V s) respectively. Simulation results have shown the effects of single and mixed carrier models, activation energies, effective mass and doping concentrations. These studies have been helpful in the development of single crystal diamond for diode applications. Reference materials of Ge and GaAs were used to test the Hall Effect system. The system was also used to characterize polycrystalline diamond deposited on glass for electrochemical applications, and Mg2(Si,Sn) compounds which are promising candidates of low-cost, light weight and non

  1. Design and realization of temperature measurement system based on optical fiber temperature sensor for wireless power transfer

    Science.gov (United States)

    Chen, Xi; Zeng, Shuang; Liu, Xiulan; Jin, Yuan; Li, Xianglong; Wang, Xiaochen

    2018-02-01

    The electric vehicles (EV) have become accepted by increasing numbers of people for the environmental-friendly advantages. A novel way to charge the electric vehicles is through wireless power transfer (WPT). The wireless power transfer is a high power transfer system. The high currents flowing through the transmitter and receiver coils increasing temperature affects the safety of person and charging equipment. As a result, temperature measurement for wireless power transfer is needed. In this paper, a temperature measurement system based on optical fiber temperature sensors for electric vehicle wireless power transfer is proposed. Initially, the thermal characteristics of the wireless power transfer system are studied and the advantages of optical fiber sensors are analyzed. Then the temperature measurement system based on optical fiber temperature sensor is designed. The system consists of optical subsystem, data acquisition subsystem and data processing subsystem. Finally, the system is tested and the experiment result shows that the system can realize 1°C precision and can acquire real-time temperature distribution of the coils, which can meet the requirement of the temperature measuring for wireless power transfer.

  2. Low-Cost Wireless Temperature Measurement: Design, Manufacture, and Testing of a PCB-Based Wireless Passive Temperature Sensor.

    Science.gov (United States)

    Yan, Dan; Yang, Yong; Hong, Yingping; Liang, Ting; Yao, Zong; Chen, Xiaoyong; Xiong, Jijun

    2018-02-10

    Low-cost wireless temperature measurement has significant value in the food industry, logistics, agriculture, portable medical equipment, intelligent wireless health monitoring, and many areas in everyday life. A wireless passive temperature sensor based on PCB (Printed Circuit Board) materials is reported in this paper. The advantages of the sensor include simple mechanical structure, convenient processing, low-cost, and easiness in integration. The temperature-sensitive structure of the sensor is a dielectric-loaded resonant cavity, consisting of the PCB substrate. The sensitive structure also integrates a patch antenna for the transmission of temperature signals. The temperature sensing mechanism of the sensor is the dielectric constant of the PCB substrate changes with temperature, which causes the resonant frequency variation of the resonator. Then the temperature can be measured by detecting the changes in the sensor's working frequency. The PCB-based wireless passive temperature sensor prototype is prepared through theoretical design, parameter analysis, software simulation, and experimental testing. The high- and low-temperature sensing performance of the sensor is tested, respectively. The resonant frequency decreases from 2.434 GHz to 2.379 GHz as the temperature increases from -40 °C to 125 °C. The fitting curve proves that the experimental data have good linearity. Three repetitive tests proved that the sensor possess well repeatability. The average sensitivity is 347.45 KHz / ℃ from repetitive measurements conducted three times. This study demonstrates the feasibility of the PCB-based wireless passive sensor, which provides a low-cost temperature sensing solution for everyday life, modern agriculture, thriving intelligent health devices, and so on, and also enriches PCB product lines and applications.

  3. Application of new design methodologies to very high-temperature metallic components of the HTTR

    International Nuclear Information System (INIS)

    Hada, Kazuhiko; Ohkubo, Minoru; Baba, Osamu

    1991-01-01

    The high-temperature piping and helium-to-helium intermediate heat exchanger of the High-Temperature Engineering Test Reactor (HTTR) are designed to be operating at very high temperatures of about 900deg C among the class 1 components of the HTTR. At such a high temperature, mechanical strength of heat-resistant metallic materials is very low and thermal expansions of structural members are large. Therefore, innovative design methodologies are needed to reduce both mechanical and thermal loads acting on these components. To the HTTR, the design methodologies which can separate the heat-resistant function from the pressure-retaining functions and allow them to expand freely are applied to reduce pressure and thermal loads. Since these design methodologies need to verify their applicability, the Japan Atomic Energy Research Institute (JAERI) has been performing many design and research works on their verifications. The details of the design methodologies and their verifications are given in this paper. (orig.)

  4. HTGR high temperature process heat design and cost status report

    International Nuclear Information System (INIS)

    1981-12-01

    This report describes the status of the studies conducted on the 850 0 C ROT indirect cycle and the 950 0 C ROT direct cycle through the end of Fiscal Year 1981. Volume I provides summaries of the design and optimization studies and the resulting capital and product costs, for the HTGR/thermochemical pipeline concept. Additionally, preliminary evaluations are presented for coupling of candidate process applications to the HTGR system

  5. Elevated temperature design of KALIMER reactor internals accounting for creep and stress-rupture effects

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Yoo, Bong

    2000-01-01

    In most LMFBR (Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER (Korea Advanced Liquid Metal Reactor) reactor internal structures is carried out for normal operating conditions which have the operating temperature 530 deg. C and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME code case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects. (author)

  6. A sandwich-designed temperature-gradient incubator for studies of microbial temperature responses

    DEFF Research Database (Denmark)

    Elsgaard, Lars; Jørgensen, L.W.

    2002-01-01

    of 28-ml test tubes. An electric plate heats one end of the TGI end and the other end is cooled by thermoelectric Peltier elements in combination with a liquid cooling system. The TGI is equipped with 24 calibrated Pt-100 temperature sensors and insulated by polyurethane plates. A PC-operated SCADA...

  7. Preconceptual design of hyfire. A fusion driven high temperature electrolysis plant

    International Nuclear Information System (INIS)

    Varljen, T.C.; Chi, J.W.H.; Karbowski, J.S.

    1983-01-01

    Brookhaven National Laboratory has been engaged in a scoping study to investigate the potential merits of coupling a fusion reactor with a high temperature blanket to a high temperature electrolysis (HTE) process to produce hydrogen and oxygen. Westinghouse is assisting this study in the areas of systems design integration, plasma engineering, balance of plant design and electrolyzer technology. The aim of the work done in the past year has been to focus on a reference design point for the plant, which has been designated HYFIRE. In prior work, the STARFIRE commercial tokamak fusion reactor was directly used as the fusion driver. This report describes a new design obtained by scaling the basic STARFIRE design to permit the achievement of a blanket power of 6000 MWt. The high temperature blanket design employs a thermally insulated refractory oxide region which provides high temperature (>1000 deg. C) steam at moderate pressures to high temperature electrolysis units. The electrolysis process selected is based on the high temperature, solid electrolyte fuel cell technology developed by Westinghouse. An initial process design and plant layout has been completed; component cost and plant economics studies are now underway to develop estimates of hydrogen production costs and to determine the sensitivity of this cost to changes in major design parameters. (author)

  8. A design method to isothermalize the core of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Takano, M.; Sawa, K.

    1987-01-01

    A practical design method is developed to isothermalize the core of block-type high-temperature gas-cooled reactors (HTGRs). Isothermalization plays an important role in increasing the design margin on fuel temperature. In this method, the fuel enrichment and the size and boron content of the burnable poison rod are determined over the core blockwise so that the axially exponential and radially flat power distribution are kept from the beginning to the end of core life. The method enables conventional HTGRs to raise the outlet gas temperature without increasing the maximum fuel temperature

  9. Rational design of temperature swing adsorption cycles for post-combustion CO2 capture

    NARCIS (Netherlands)

    Joss, Lisa; Gazzani, Matteo; Mazzotti, Marco

    2017-01-01

    The design of temperature swing adsorption (TSA) cycles aimed at recovering the heavy product at high purity is investigated by model-based design and applied to the capture of CO2 from flue gases. This model based design strategy and an extensive parametric analysis enables gaining an understanding

  10. Ambient Temperature Based Thermal Aware Energy Efficient ROM Design on FPGA

    DEFF Research Database (Denmark)

    Saini, Rishita; Bansal, Neha; Bansal, Meenakshi

    2015-01-01

    Thermal aware design is currently gaining importance in VLSI research domain. In this work, we are going to design thermal aware energy efficient ROM on Virtex-5 FPGA. Ambient Temperature, airflow, and heat sink profile play a significant role in thermal aware hardware design life cycle. Ambient...

  11. Designing electronic equipment on the basis of standard mechanical structures using internet re­sour­ces

    Directory of Open Access Journals (Sweden)

    Karlangach A. P.

    2016-12-01

    Full Text Available The author proposes a method to design electronic equipment based on functional-node design method that involves the use of 2D- and 3D- models mechanical structures for electronic equipment as a way to reduce development time and errors when creating design documentation for electronic equipment. At present, most areas of science and technology are computerized, more problems in designing electronic equipment are dealt with using computer-aided design (CAD and Computer-aided manufacturing (CAM to reduce the time required for development and manufacturing of electronic equipment. Development of design documentation also requires a more effective approach, because the less the time for development of the design documentation is, the faster the developed device will go into production. The aim of the study is to develop a method of designing electronic equipment using 2D and 3D models of standard mechanical structures for electronic equipment using Internet resources. Based on the presented methods is an example of designing a device from standard bearing structures. Compared with traditional technology, the method of designing electronic equipment using standard parts has the following advantages: - reduces time and improves quality of development through the use of existing design documentation; - accelerates the implementation and introducing into production processes; - increases unification of design solutions.

  12. Off-design performance analysis of Kalina cycle for low temperature geothermal source

    International Nuclear Information System (INIS)

    Li, Hang; Hu, Dongshuai; Wang, Mingkun; Dai, Yiping

    2016-01-01

    Highlights: • The off-design performance analysis of Kalina cycle is conducted. • The off-design models are established. • The genetic algorithm is used in the design phase. • The sliding pressure control strategy is applied. - Abstract: Low temperature geothermal sources with brilliant prospects have attracted more and more people’s attention. Kalina cycle system using ammonia water as working fluid could exploit geothermal energy effectively. In this paper, the quantitative analysis of off-design performance of Kalina cycle for the low temperature geothermal source is conducted. The off-design models including turbine, pump and heat exchangers are established preliminarily. Genetic algorithm is used to maximize the net power output and determine the thermodynamic parameters in the design phase. The sliding pressure control strategy applied widely in existing Rankine cycle power plants is adopted to response to the variations of geothermal source mass flow rate ratio (70–120%), geothermal source temperature (116–128 °C) and heat sink temperature (0–35 °C). In the off-design research scopes, the guidance for pump rotational speed adjustment is listed to provide some reference for off-design operation of geothermal power plants. The required adjustment rate of pump rotational speed is more sensitive to per unit geothermal source temperature than per unit heat sink temperature. Influence of the heat sink variation is greater than that of the geothermal source variation on the ranges of net power output and thermal efficiency.

  13. Design and evaluation of a pressure sensor for high temperature nuclear application

    International Nuclear Information System (INIS)

    Yancey, M.E.

    1981-11-01

    The goal of this technical development task was the development of a small eddy-current pressure sensor for use within a high temperature nuclear environment. The sensor is designed for use at pressures and temperatures of up to 17.23 MPa and 650 0 F. The design of the sensor incorporated features to minimize possible errors due to temperature transients present in nuclear applications. This report describes a prototype pressure sensor that was designed, the associated 100 kHz signal conditioning electronics, and the evaluation tests which were conducted

  14. Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: Conclusions from an experimental program in a 4-loop test facility (PKL)

    International Nuclear Information System (INIS)

    Umminger, K.J.; Kastner, W.; Mandl, R.M.; Weber, P.

    1993-01-01

    Within the scope of German reactor safety research, extensive experiments covering the behavior of nuclear power plants under accident conditions have been carried out in the PKL test facility which simulates a 4-loop, 1,300 MWe KWU-designed PWR. While the investigations dealing with design-basis accidents and with the efficiency of the emergency core cooling systems have been largely completed, the main interest nowadays concentrates on the investigation of beyond-design-basis accidents to demonstrate the safety margins of nuclear power plants and to investigate the contribution of the built-in safety features for a further reduction of the residual risk. The thermal hydraulic behavior of a PWR under these extreme accident conditions was experimentally investigated within the PKL III B test program. This paper presents the fundamental findings with some of the most important results being discussed in detail. Future plans are also outlined

  15. Shape evolution of new-phased lepidocrocite VOOH from single-shelled to double-shelled hollow nanospheres on the basis of programmed reaction-temperature strategy.

    Science.gov (United States)

    Wu, Changzheng; Zhang, Xiaodong; Ning, Bo; Yang, Jinlong; Xie, Yi

    2009-07-06

    Solid templates have been long regarded as one of the most promising ways to achieve single-shelled hollow nanostructures; however, few effective methods for the construction of multishelled hollow objects from their solid template counterparts have been developed. We report here, for the first time, a novel and convenient route to synthesizing double-shelled hollow spheres from the solid templates via programming the reaction-temperature procedures. The programmed temperature strategy developed in this work then provides an essential and general access to multishelled hollow nanostructures based on the designed extension of single-shelled hollow objects, independent of their outside contours, such as tubes, hollow spheres, and cubes. Starting from the V(OH)(2)NH(2) solid templates, we show that the relationship between the hollowing rate and the reaction temperature obey the Van't Hoff rule and Arrhenius activation-energy equation, revealing that it is the chemical reaction rather than the diffusion process that guided the whole hollowing process, despite the fact that the coupled reaction/diffusion process is involved in the hollowing process. Using the double-shelled hollow spheres as the PCM (CaCl(2).6H(2)O) matrix grants much better thermal-storage stability than that for the nanoparticles counterpart, revealing that the designed nanostructures can give rise to significant improvements for the energy-saving performance in future "smart house" systems.

  16. Design of high precision temperature control system for TO packaged LD

    Science.gov (United States)

    Liang, Enji; Luo, Baoke; Zhuang, Bin; He, Zhengquan

    2017-10-01

    Temperature is an important factor affecting the performance of TO package LD. In order to ensure the safe and stable operation of LD, a temperature control circuit for LD based on PID technology is designed. The MAX1978 and an external PID circuit are used to form a control circuit that drives the thermoelectric cooler (TEC) to achieve control of temperature and the external load can be changed. The system circuit has low power consumption, high integration and high precision,and the circuit can achieve precise control of the LD temperature. Experiment results show that the circuit can achieve effective and stable control of the laser temperature.

  17. Analysis and evaluation system for elevated temperature design of pressure vessels

    International Nuclear Information System (INIS)

    Hayakawa, Teiji; Sayawaki, Masaaki; Nishitani, Masahiro; Mii, Tatsuo; Murasawa, Kanji

    1977-01-01

    In pressure vessel technology, intensive efforts have recently been made to develop the elevated temperature design methods. Much of the impetus of these efforts has been provided mainly by the results of the Liquid Metal Fast Breeder Reactor (LMFBR) and more recently, of the High Temperature Gas-cooled Reactor (HTGR) Programs. The pressure vessels and associated components in these new type nuclear power plants must operate for long periods at elevated temperature where creep effects are significant and then must be designed by rigorous analysis for high reliability and safety. To carry out such an elevated temperature designing, numbers of highly developed analysis and evaluation techniques, which are so complicated as to be impossible by manual work, are indispensable. Under these circumstances, the authors have made the following approaches in the study: (1) Study into basic concepts and the associated techniques in elevated temperature design. (2) Systematization (Analysis System) of the procedure for loads and stress analyses. (3) Development of post-processor, ''POST-1592'', for strength evaluation based on ASME Code Case 1592-7. By linking the POST-1592 together with the Analysis System, an analysis and evaluation system is developed for an elevated temperature design of pressure vessels. Consequently, designing of elevated temperature vessels by detailed analysis and evaluation has easily and effectively become feasible by applying this software system. (auth.)

  18. High temperature structural design and R and Ds for heat transport system components of FBR 'Monju'

    International Nuclear Information System (INIS)

    Sumikawa, Masaharu; Nakagawa, Yukio; Fukuda, Yoshio; Sukegawa, Masayuki; Ishizaki, Tairo.

    1980-01-01

    The machines and equipments of cooling system for the fast breeder prototype reactor ''Monju'' are operated in creep temperature region, and the upper limit temperature to apply the domestic structural design standard for nuclear machines and equipment is exceeded, therefore the guideline for high temperature structural design is being drawn up, reflecting the results of recent research and development, by the Power Reactor and Nuclear Fuel Development Corp. and others. In order to obtain the basic data for the purpose, the tests on the high temperature characteristics of main structural members and structural elements were carried out, and eight kinds of the inelastic structural analysis program ''HI-EPIC'' series were developed, thus the fundamental technologies of structural desigh in non-linear region were established. Also in the non-linear region, enormous physical quantities must be evaluated, and in the design method based on real elastic analysis, many design diagrams must be employed, therefore for the purpose of improving the reliability of evaluation, the automatic evaluation program ''HI-TEP'' was developed, and preparation has been made for the design of actual machines. The high temperature structural design in ''Monju'', the development of inelastic structural analysis program and high temperature structural analysis evaluation program, and the development of high temperature structures and materials are described. (Kako, I.)

  19. Operating experience and systems analysis at Trillo NPP: A program intended for systematic review of plant safety systems to assess design basis requirements compliance

    International Nuclear Information System (INIS)

    Vega, R. de la

    1996-01-01

    The program was defined to apply to all plant safety systems and/or systems included in plant Technical Specifications. The goal of the program was to ensure, by systematic design, construction, and commissioning review, the adequacy of safety systems, structures and components to fulfill their safety functions. Also, as a result of the program, it was established that a complete, unambiguous, systematic, design basis definition shall take place. And finally, a complete documental review of the plant design shall result from the program execution

  20. [The design of heat dissipation of the field low temperature box for storage and transportation].

    Science.gov (United States)

    Wei, Jiancang; Suin, Jianjun; Wu, Jian

    2013-02-01

    Because of the compact structure of the field low temperature box for storage and transportation, which is due to the same small space where the compressor, the condenser, the control circuit, the battery and the power supply device are all placed in, the design for heat dissipation and ventilation is of critical importance for the stability and reliability of the box. Several design schemes of the heat dissipation design of the box were simulated using the FLOEFD hot fluid analysis software in this study. Different distributions of the temperature field in every design scheme were constructed intimately in the present study. It is well concluded that according to the result of the simulation analysis, the optimal heat dissipation design is decent for the field low temperature box for storage and transportation, and the box can operate smoothly for a long time using the results of the design.

  1. Technical basis for the reduction of the maximum temperature TGA-MS analysis of oxide samples from the 3013 destructive examination program

    International Nuclear Information System (INIS)

    Scogin, J. H.

    2016-01-01

    Thermogravimetric analysis with mass spectroscopy of the evolved gas (TGA-MS) is used to quantify the moisture content of materials in the 3013 destructive examination (3013 DE) surveillance program. Salts frequently present in the 3013 DE materials volatilize in the TGA and condense in the gas lines just outside the TGA furnace. The buildup of condensate can restrict the flow of purge gas and affect both the TGA operations and the mass spectrometer calibration. Removal of the condensed salts requires frequent maintenance and subsequent calibration runs to keep the moisture measurements by mass spectroscopy within acceptable limits, creating delays in processing samples. In this report, the feasibility of determining the total moisture from TGA-MS measurements at a lower temperature is investigated. A temperature of the TGA-MS analysis which reduces the complications caused by the condensation of volatile materials is determined. Analysis shows that an excellent prediction of the presently measured total moisture value can be made using only the data generated up to 700 °C and there is a sound physical basis for this estimate. It is recommended that the maximum temperature of the TGA-MS determination of total moisture for the 3013 DE program be reduced from 1000 °C to 700 °C. It is also suggested that cumulative moisture measurements at 550 °C and 700°C be substituted for the measured value of total moisture in the 3013 DE database. Using these raw values, any of predictions of the total moisture discussed in this report can be made.

  2. Technical basis for the reduction of the maximum temperature TGA-MS analysis of oxide samples from the 3013 destructive examination program

    Energy Technology Data Exchange (ETDEWEB)

    Scogin, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-24

    Thermogravimetric analysis with mass spectroscopy of the evolved gas (TGA-MS) is used to quantify the moisture content of materials in the 3013 destructive examination (3013 DE) surveillance program. Salts frequently present in the 3013 DE materials volatilize in the TGA and condense in the gas lines just outside the TGA furnace. The buildup of condensate can restrict the flow of purge gas and affect both the TGA operations and the mass spectrometer calibration. Removal of the condensed salts requires frequent maintenance and subsequent calibration runs to keep the moisture measurements by mass spectroscopy within acceptable limits, creating delays in processing samples. In this report, the feasibility of determining the total moisture from TGA-MS measurements at a lower temperature is investigated. A temperature of the TGA-MS analysis which reduces the complications caused by the condensation of volatile materials is determined. Analysis shows that an excellent prediction of the presently measured total moisture value can be made using only the data generated up to 700 °C and there is a sound physical basis for this estimate. It is recommended that the maximum temperature of the TGA-MS determination of total moisture for the 3013 DE program be reduced from 1000 °C to 700 °C. It is also suggested that cumulative moisture measurements at 550 °C and 700°C be substituted for the measured value of total moisture in the 3013 DE database. Using these raw values, any of predictions of the total moisture discussed in this report can be made.

  3. Design and development of gas turbine high temperature reactor 300 (GTHTR300)

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Takizuka, Takakazu; Yan, Xing; Kosugiyama, Shinichi

    2003-01-01

    JAERI (Japan Atomic Energy Research Institute) started design and development of the high temperature gas cooled reactor with a gas turbine electric generation system, GTHTR300, in April 2001. Design originalities of the GTHTR300 are a horizontally mounted highly efficient gas turbine system and an ultimately simplified safety system such as no containment building and no active emergency core cooling. These design originalities are proposed based on design and operational experiences in conventional gas turbine systems and Japan's first high temperature gas cooled reactor (HTTR: High Temperature Engineering Test Reactor) so that many R and Ds are not required for the development. Except these original design features, devised core design, fuel design and plant design are adopted to meet design requirements and attain a target cost. This paper describes the unique design features focusing on the safety design, reactor core design and gas turbine system design together with a preliminary result of the safety evaluation carried out for a typical severe event. This study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  4. Effect of design factors on surface temperature and wear in disk brakes

    Science.gov (United States)

    Santini, J. J.; Kennedy, F. E.; Ling, F. F.

    1976-01-01

    The temperatures, friction, wear and contact conditions that occur in high energy disk brakes are studied. Surface and near surface temperatures were monitored at various locations in a caliper disk brake during drag type testing, with friction coefficient and wear rates also being determined. The recorded transient temperature distributions in the friction pads and infrared photographs of the rotor disk surface both showed that contact at the friction surface was not uniform, with contact areas constantly shifting due to nonuniform thermal expansion and wear. The effect of external cooling and of design modifications on friction, wear and temperatures was also investigated. It was found that significant decreases in surface temperature and in wear rate can be achieved without a reduction in friction either by slotting the contacting face of the brake pad or by modifying the design of the pad support to improve pad compliance. Both design changes result in more uniform contact conditions on the friction surface.

  5. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  6. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  7. Design rules for high temperature plant - the implications of recent research in relation to current practice

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1977-01-01

    An historical summary is presented of design rules for high temperature plant, leading to the rules applicable to high temperature reactors, particularly the liquid metal fast breeder reactor. Special attention is given to creep rupture properties of ferritic and austenitic materials used for the construction of components such as boilers and pressure vessels. (author)

  8. Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lee, Won-Jae

    2006-01-01

    The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

  9. Research Activities on Development of Piping Design Methodology of High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Nam-Su [Seoul National Univ. of Science and Technology, Seoul(Korea, Republic of); Won, Min-Gu [Sungkyukwan Univ., Suwon (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering and Construction Co. Inc., Gimcheon (Korea, Republic of); Lee, Hyeog-Yeon; Kim, Yoo-Gon [Korea Atomic Energy Research Institute, Daejeon(Korea, Republic of)

    2016-10-15

    A SFR is operated at high temperature and low pressure compared with commercial pressurized water reactor (PWR), and such an operating condition leads to time-dependent damages such as creep rupture, excessive creep deformation, creep-fatigue interaction and creep crack growth. Thus, high temperature design and structural integrity assessment methodology should be developed considering such failure mechanisms. In terms of design of mechanical components of SFR, ASME B and PV Code, Sec. III, Div. 5 and RCC-MRx provide high temperature design and assessment procedures for nuclear structural components operated at high temperature, and a Leak-Before-Break (LBB) assessment procedure for high temperature piping is also provided in RCC-MRx, A16. Three web-based evaluation programs based on the current high temperature codes were developed for structural components of high temperature reactors. Moreover, for the detailed LBB analyses of high temperature piping, new engineering methods for predicting creep C*-integral and creep COD rate based either on GE/EPRI or on reference stress concepts were proposed. Finally, the numerical methods based on Garofalo's model and RCC-MRx have been developed, and they have been implemented into ABAQUS. The predictions based on both models were compared with the experimental results, and it has been revealed that the predictions from Garafalo's model gave somewhat successful results to describe the deformation behavior of Gr. 91 at elevated temperatures.

  10. Regulatory requirements on accident management and emergency preparedness - concept of nuclear and radiation safety during beyond-design-basis accidents

    International Nuclear Information System (INIS)

    Yanke, R.

    2002-01-01

    Actual practice the and proposals for further activities in the field of Accident Management (AM) in the member countries of the Co-operation Forum of WWER regulators and in Western countries have been assessed. Further the results of the last working group on AM , the overview of interactions of severe accident research and the regulatory positions in various countries, IAEA reports, practice in Switzerland and Finland, were taken into consideration. From this information, the working group derived recommendations on Accident Management. The general proposals correspond to the present state of the art on AM. They do not describe the whole spectra of recommendations on AM for NPPs with WWER reactors. A basis for the implementation of an AM program is given, which could be extended in a follow-up working group. The developments and research concerning AM have to be continued. The positions of various countries with regard to the 'Interactions of severe accident research and the regulatory positions' are given. On the basis of the working group proposals, the WWER regulators could set regulatory requirements and support further developments of AM strategies, making use of the benefits of common features of NPPs with WWER reactors. Concerted actions in the field of AM between the WWER regulators would bundle the development of a unified concept of recommendations and speed up the implementation of AM measures in order to minimise the risks involved in nuclear power generation

  11. The PLC-based Industrial Temperature Control System: Design and Implementation

    Directory of Open Access Journals (Sweden)

    Wei Fanjie

    2017-01-01

    Full Text Available Targeting at the problem of slow response and low accuracy of the automatic temperature control system for material processing and boiler heating, a new design method is proposed to work with the PLC-based temperature control system, where the box temperature control may be achieved through the fan and the heating plate. The hardware design and software design of the system are analyzed in detail. In this paper, a combination of the traditional PID control and the more popular fuzzy control is taken as the control program to achieve the overall design of the control algorithm. Followed by the simulation in the MATLAB software, the designed system is highlighted by its the characteristics of impressive stability, precision and robustness.

  12. Design of laser diode driver with constant current and temperature control system

    Science.gov (United States)

    Wang, Ming-cai; Yang, Kai-yong; Wang, Zhi-guo; Fan, Zhen-fang

    2017-10-01

    A laser Diode (LD) driver with constant current and temperature control system is designed according to the LD working characteristics. We deeply researched the protection circuit and temperature control circuit based on thermos-electric cooler(TEC) cooling circuit and PID algorithm. The driver could realize constant current output and achieve stable temperature control of LD. Real-time feedback control method was adopted in the temperature control system to make LD work on its best temperature point. The output power variety and output wavelength shift of LD caused by current and temperature instability were decreased. Furthermore, the driving current and working temperature is adjustable according to specific requirements. The experiment result showed that the developed LD driver meets the characteristics of LD.

  13. Assessment of value-impact associated with the elimination of postulated pipe ruptures from the design basis for nuclear power plants

    International Nuclear Information System (INIS)

    Holman, G.S.; Chou, C.K.

    1985-01-01

    The US Nuclear Regulatory Commission is proposing to amend the regulations that currently require that the design basis for nuclear power plants include the postulation of dynamic effects from loss of coolant accidents up to and including the double-ended rupture of the largest pipe in the reactor coolant system. Proposed modifications would allow analyses to serve as a sufficient basis for excluding dynamic effects, including but not necessarily limited to pipe whip and jet impingement, associated with specific pipe ruptures. Only dynamic effects would be impacted; current design requirements for containment sizing and discharge capacity of emergency core cooling systems would remain unchanged. This report presents a detailed analysis of value-impact associated with the proposed amendment for PWR reactor coolant loop piping and for BWR recirculation loop piping. The effect of extending application of the proposed rule change to other piping systems is also assessed in a less quantitative manner

  14. Null testing of nonrotational symmetry transmission optical freeform: design, modeling, and inspection on the basis of Fermat principles

    Science.gov (United States)

    Qiu, Gufeng; Cui, Xudong

    2015-11-01

    We present a general design method for a type of transmission freeforms without rotational symmetry and achieve the null testing by putting a well-designed Fermat reflector on the transmitting optical path. The design principle of the reflector is given, and an eccentric spherical surface with 1-mm deviation is used as an example of testing freeform. We fabricated the reflector and the freeform with the single-point diamond turning machine. Both conventional interference inspection and our approach give consistent results. The design error is less than 106 mm, and the measurement accuracy is nearly completely determined by the fabrication precision. This approach can also be applied to the inspections of reflecting freeforms with low costs.

  15. Influence of the design temperature on long-term safety of a salt dome repository

    International Nuclear Information System (INIS)

    Buhmann, D.; Brenner, J.; Storck, R.

    1993-03-01

    All studies made so far within the framwork of the mixed concept system analysis proceeded from a design temperature of the mine structure of 200 C. The concept based on a design temperature of 150 C was aimed at studying whether it made sense to maintain lower temperatures, if necessary. Deterministic and probabilistic calculations were made in order to determine the influence of the lower design temperature on long-term safety. The calculations were based on concept A of Joint Borehole and Gallery Storage. Assuming reference values of the input parameters, the deterministic calculations do not produce any radionuclide release from the mine structure. If, however, one assumes a lower rate for rock convergence, radionuclides are released at maximum dose rates of about 3.10 -5 Sv/a. Even a larger volume of limited brine inclusions may lead to radionuclide releases, in that case with dose commitments of the order of magnitude of 1.10 -5 Sv/a. The probabilistic calculations show that a design temperature of 150 C for long-term safety is less favourable than a higher design temperature. The share of simulations in the probabilistic calculations with a radionuclide release, and the expected value of dose commitment, are almost double as high as in the concept based on 200 C design temperature. Thus a higher design temperature is preferable with regard to the long-term safety of a salt repository. The most important parameters concerning dose commitment are the volume of limited brine inclusions, the convergence rate, and the permeability of barriers and backfilling rock. (orig./HP) [de

  16. Regulatory standpoints on the design-basis capability of safety-related motor-operated valves(MOVs) and power-operated gate valves(POGVs)

    International Nuclear Information System (INIS)

    Kim, W. T.; Kum, O. H.

    1999-01-01

    The weakness in the design-basis capability of Motor-Operated Valves(MOVs) and the susceptibility to Pressure Locking and Thermal Binding phenomena of Power-Operated Gate Valves(POGVs) have been major concerns to be resolved in the nuclear society in and abroad since Three Mile Island accident occurred in the USA in 1979. Through detailed analysis of operating experience and regulatory activities, some MOVs and POGVs have been found to be unreliable in performing their safety functions when they are required to do so under certain conditions, especially under design-basis accident conditions. Further, it is well understood that these safety problems may not be identified by the typical valve in-service testing(IST). USNRC has published three Generic Letters, GL 89-10, GL 95-07, and GL 96-05, requiring nuclear plant licensees to take appropriate actions to resolve the problems mentioned above. Korean nuclear regulatory body has made public an administration measure called 'Regulatory recommendation to verify safety functions of the safety-related MOVs and POGVs' on June 13, 1997, and in this administration measure Korean utility is asked to submit written documents to show how it assure design-basis capability of these valves. The following are among the major concerns being considered from a regulation standpoint. Program scope and implementation priority, dynamic tests under differential pressure conditions, accuracy of diagnostic equipment, torque switch setting and torque bypass percentage, weak link analysis, motor actuator sizing, corrective actions taken to resolve pressure locking and thermal binding susceptibility, and a periodic verification program for the valves once design-basis capability has been verified

  17. High-Temperature Air-Cooled Power Electronics Thermal Design: Annual Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Waye, Scot [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-08-01

    Power electronics that use high-temperature devices pose a challenge for thermal management. With the devices running at higher temperatures and having a smaller footprint, the heat fluxes increase from previous power electronic designs. This project overview presents an approach to examine and design thermal management strategies through cooling technologies to keep devices within temperature limits, dissipate the heat generated by the devices and protect electrical interconnects and other components for inverter, converter, and charger applications. This analysis, validation, and demonstration intends to take a multi-scale approach over the device, module, and system levels to reduce size, weight, and cost.

  18. Numerical studies of large penetrations and closures for containment vessels subjected to loadings beyond the design basis

    International Nuclear Information System (INIS)

    Kulak, R.F.; Hsieh, B.J.; Kennedy, J.M.; Ash, J.E.; McLennan, G.A.

    1984-01-01

    Numerical simulations of the macro-deformations of the sealing surfaces (gasketed junctures) of a PWR steel containment vessel's equipment hatch and a BWR Mk II containment vessel head have been performed. Results for the equipment hatch juncture indicate that the rotations of the hatch cover and penetration sleeve must be accounted for when performing leakage analysis because they can effect the compression of the gasket even though the gasket is in a pressure-seated configuration. Results from a leakage analysis indicated that excessive leakage can occur if the surface roughness is high and/or the compression set is high. Results for the Mk II head show that both the temperature and pressure loadings must be taken into account to obtain realistic responses. The temperature difference between the flanges and bolts has the important net effect of keeping the gasketed juncture closed, that is in metal-to-metal contact. Due to the high accident temperature, the gasket itself was found to achieve 100% compression set and thus could not perform its sealing function within the juncture

  19. Characteristic features of the core design of high-temperature reactors

    International Nuclear Information System (INIS)

    Brandes, S.; Lohnert, G.

    1975-01-01

    Following a survey on the possible applications of the HTGR depending on the height of the gas exiting temperatures, the core design for both of the fuel element concepts 'sphere' and 'block' is dealt with. The particularities arising from the multiple refueling and the one-way fueling in the design for spherical fuel elements are discussed. (UA/LH) [de

  20. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.

  1. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs

  2. Design of a high-temperature superconductor current lead for electric utility SMES

    International Nuclear Information System (INIS)

    Niemann, R.C.; Cha, Y.S.; Hull, J.R.; Rey, C.M.; Dixon, K.D.

    1995-01-01

    Current leads that rely on high-temperature superconductors (HTSs) to deliver power to devices operating at liquid helium temperature have the potential to reduce refrigeration requirements to levels significantly below those achievable with conventional leads. The design of HTS current leads suitable for use in near-term superconducting magnetic energy storage (SMES) is in progress. The SMES system has an 0.5 MWh energy capacity and a discharge power of 30 MW. Lead-design considerations include safety and reliability, electrical and thermal performance, structural integrity, manufacturability, and cost. Available details of the design, including materials, configuration, and performance predictions, are presented

  3. Design of the steam reformer for the HTR-10 high temperature process heat application

    International Nuclear Information System (INIS)

    Ju Huaiming; Xu Yuanhui; Jia Haijun

    2000-01-01

    The 10 MW High Temperature Reactor Test Module (HTR-10) is being constructed now and planned to be operational in 2000. One of the objectives is to develop the high temperature process heat application. The methane steam reformer is one of the key-facilities for the nuclear process heat application system. The paper describes the conceptual design of the HTR-10 Steam Reformer with He heating, and the design optimization computer code. It can be used to perform sensitivity analysis for parameters, and to improve the design. Principal parameters and construction features of the HTR-10 reformer heated by He are introduced. (author)

  4. Requirements on the mechanical design of reactor systems operating at elevated temperature

    International Nuclear Information System (INIS)

    Schulz, H.; Glahn, M.

    1979-01-01

    The paper presents the contemporary status of the requirements on the mechanical design and analysis developed during the licensing procedure of reactor systems operating at elevated temperature. General requirements for the design at elevated temperature are reviewed. The main proposal is to point out some limit strain criteria which are not included in present design guidelines and codes. The developed strain criteria are used to limit the component deformations in case of power excursions like the Bethe-Tait accident. It is also applicable for loads arising from other faulted conditions. (orig.)

  5. Interrelationship betwen material strength and component design under elevated temperature for FBR

    International Nuclear Information System (INIS)

    Nakagawa, Y.

    Structural design under elevated temperature for fast breeder reactor plant is very troublesome compared to that of for lower temperature. This difficulty can be mainly discussed from two different stand points. One is design and design code, another is material strength. Components in FBR are operated under creep regime and time dependent creep behaviour should be elevated properly. This means the number and combinations of design code and material strength are significantly large and makes these systems very complicated. Material selection is, in no words, not an easy job. This should be done by not only material development but also component design stand point. With valuable experience of construction and research on FBR, a lot of information on component design and material behaviour is available. And it is a time to choose the ''best material'' from the entire stand points of component construction. (author)

  6. Designing and Evaluating Tutoring Feedback Strategies for Digital Learning Environments on the Basis of the Interactive Tutoring Feedback Model

    Science.gov (United States)

    Narciss, Susanne

    2013-01-01

    This paper describes the interactive tutoring feedback model (ITF-model; Narciss, 2006; 2008), and how it can be applied to the design and evaluation of feedback strategies for digital learning environments. The ITF-model conceptualizes formative tutoring feedback as a multidimensional instructional activity that aims at contributing to the…

  7. Simplified methods and application to preliminary design of piping for elevated temperature service

    International Nuclear Information System (INIS)

    Severud, L.K.

    1975-01-01

    A number of simplified stress analysis methods and procedures that have been used on the FFTF project for preliminary design of piping operating at elevated temperatures are described. The rationale and considerations involved in developing the procedures and preliminary design guidelines are given. Applications of the simplified methods to a few FFTF pipelines are described and the success of these guidelines are measured by means of comparisons to pipeline designs that have had detailed Code type stress analyses. (U.S.)

  8. Task to Training Matrix Design for Decommissioning Engineer on the basis of Systematic Approach to Training Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Jeong Keun [KHNP, Ulsan (Korea, Republic of)

    2016-10-15

    In nuclear history, before Chernobyl Accident, Three Mile Island (TMI) Accident was the severest accident. For this reason, to resolve the disclosed or potential possibilities of nuclear accident, more than one hundred countermeasures were proposed by United States Nuclear Regulatory Commission (USNRC). Among various recommendations by USNRC, one suggestion was related to training aspect. It was Systematic Approach to Training (SAT) and this event was the initiation of SAT methodology in the world. In Korea, upcoming June 2017, Kori Unit-1 NPP is scheduled to be shut down and it will experience NPP decommissioning for the first time. Present study aims to establish concrete training foundation for NPP decommissioning engineers based on Systematic Approach to Training (SAT) methodology, in particular, Task to Training Matrix (TTM). The objective of this paper is to organize TTM on the basis of SAT for NPP decommissioning engineer. For this reason, eighteen tasks are yielded through Job and Task Analysis (JTA) process. After that, for the settlement of Task to Training Matrix (TTM), various data are determined such as element, condition, standard, knowledge and skill, learning objective and training setting. When it comes to training in nuclear industry, SAT methodology has been the unwavering principle in Korea since NPPs export to UAE.

  9. Optimized Design of the SGA-WZ Strapdown Airborne Gravimeter Temperature Control System

    Directory of Open Access Journals (Sweden)

    Juliang Cao

    2015-12-01

    Full Text Available The temperature control system is one of the most important subsystems of the strapdown airborne gravimeter. Because the quartz flexible accelerometer based on springy support technology is the core sensor in the strapdown airborne gravimeter and the magnet steel in the electromagnetic force equilibrium circuits of the quartz flexible accelerometer is greatly affected by temperature, in order to guarantee the temperature control precision and minimize the effect of temperature on the gravimeter, the SGA-WZ temperature control system adopts a three-level control method. Based on the design experience of the SGA-WZ-01, the SGA-WZ-02 temperature control system came out with a further optimized design. In 1st level temperature control, thermoelectric cooler is used to conquer temperature change caused by hot weather. The experiments show that the optimized stability of 1st level temperature control is about 0.1 °C and the max cool down capability is about 10 °C. The temperature field is analyzed in the 2nd and 3rd level temperature control using the finite element analysis software ANSYS. The 2nd and 3rd level temperature control optimization scheme is based on the foundation of heat analysis. The experimental results show that static accuracy of SGA-WZ-02 reaches 0.21 mGal/24 h, with internal accuracy being 0.743 mGal/4.8 km and external accuracy being 0.37 mGal/4.8 km compared with the result of the GT-2A, whose internal precision is superior to 1 mGal/4.8 km and all of them are better than those in SGA-WZ-01.

  10. Design Basis of Core Components and their Realization in the frame of the EPR'sTM Core Component Development

    International Nuclear Information System (INIS)

    Schebitz, Florian; Mekmouche, Abdelhalim

    2008-01-01

    Rod Cluster Control Assemblies (RCCAs), Thimble Plug Assemblies (TPAs), Primary Neutron Sources (PNS) and Secondary Neutron Sources (SNS) are essential for the operation of a Nuclear Power Plant. Different functional requirements ask for different components and geometries. Therefore three different core components are used within the primary circuit: - The RCCA, which contains the absorber materials, is used to regulate and shut down the nuclear chain reaction. Under these demanding conditions different effects are determining the lifetime of the RCCA and in particular of the control rods. Several improvements like ion-nitriding of the cladding, lengthening of the bottom end plug, helium backfilling and reduction of the absorber diameter in the bottom part, which have already been introduced with the HARMONI TM RCCA, show a real improvement in terms of lifetime. - The TPAs are used at positions without RCCAs and neutron sources to limit the by-pass flow-rate in the fuel assembly guide tubes. The advanced TPA design results from a perfect combination of French and German design experience feedback. Benefits like homogenized hydraulic flow and improved manageability in terms of handling tools show the joined experience. - The neutron sources are used to enhance the flux level when the core is sub-critical so as to facilitate the core start-up control by the neutron flux detectors. Primary and secondary neutron sources are designed in a common way with reviewed and improved methodology. As there are different ways and conditions to operate core components, several designs are available. For the EPR TM , the best methods and products have been chosen. All chosen components contribute to an optimized and safe operation of the EPR TM . (authors)

  11. Designing and Evaluating Tutoring Feedback Strategies for digital learning environments on the basis of the Interactive Tutoring Feedback Model

    Directory of Open Access Journals (Sweden)

    Susanne Narciss

    2013-06-01

    Full Text Available This paper describes the interactive tutoring feedback model (ITF-model; Narciss, 2006; 2008, and how it can be applied to the design and evaluation of feedback strategies for digital learning environments. The ITF-model conceptualizes formative tutoring feedback as a multidimensional instructional activity that aims at contributing to the regulation of a learning process in order to help learners acquire or improve the competencies needed to master learning tasks. It integrates findings from systems theory with recommendations of prior research on interactive instruction and elaborated feedback, on task analyses, on error analyses, and on tutoring techniques. Based on this multi-dimensional view of formative tutoring feedback methodological implications for designing and investigating multiple effects of feedback under multiple individual and situational conditions are described. Furthermore, the paper outlines how the implications of the ITF-model have been applied in several studies to the design and evaluation of tutoring feedback strategies for digital learning environments (e.g., Narciss, 2004; Narciss & Huth, 2006; Narciss, Schnaubert, Andres, Eichelmann, Goguadze, & Sosnovsky, 2013.

  12. Probabilistic risk assessment for salt repository conceptual design of subsurface facilities: A techical basis for Q-list determination

    International Nuclear Information System (INIS)

    Chen, C.P.; Mayberry, J.J.; Shepherd, J.; Koza, H.; Rahmani, H.; Sinsky, J.

    1987-12-01

    Subpart G ''Quality Assurance'' of 10 CFR Part 60 requires that the US Department of Energy (DOE) apply a quality assurance program to ''all systems, structures, and components important to safety'' and to ''design and characterization of barriers important to waste isolation.'' In April 1986, DOE's Office of Geologic Repositories (OGR) issued general guidance for formulating a list of such systems, structures, and components---the Q-list. This guidance called for the use of probabilistic risk assessment (PRA) techniques to identify Q-list items. In this report, PRA techniques are applied to the underground facilities and systems described in the conceptual design report for the Salt Repository Project (SRP) in Deaf Smith County, Texas. Based on probability and dose consequence calculations, no specific items were identified for the Q-list. However, evaluation of the analyses indicated that two functions are important in precluding off-site releases of radioactivity: disposal container integrity; and isolation of the underground facility by the heating, ventilation, and air conditioning (HVAC) systems. Items related to these functions are recommended for further evaluation as the repository design progresses. 13 refs., 20 figs

  13. Thermodynamic analysis of a new design of temperature controlled parabolic trough collector

    International Nuclear Information System (INIS)

    Ceylan, İlhan; Ergun, Alper

    2013-01-01

    Highlights: • This new design parabolic trough collector has been made as temperature control. • The TCPTC system is very appropriate for the industrial systems which require high temperatures. • With TCPTC can provide hot water with low solar radiation. • TCPTC system costs are cheaper than other systems (thermo siphon systems, pomp systems, etc.). - Abstract: Numerous types of solar water heater are used throughout the world. These heaters can be classified into two groups as pumped systems and thermo siphon systems. However, water temperature cannot be controlled by these systems. In this study, a new temperature-controlled parabolic trough collector (TCPTC) was designed and analyzed experimentally. The analysis was made at a temperature range of 40–100 °C, with at intervals of 10 °C. A detailed analysis was performed by calculating energy efficiencies, exergy efficiencies, water temperatures and water amounts. The highest energy efficiency of TCPTC was calculated as 61.2 for 100 °C. As the set temperature increased, the energy efficiency increased as well. The highest exergy efficiency was calculated as 63 for 70 °C. However, as the set temperature increased, the exergy efficiency did not increase. Optimum exergy efficiency was obtained for 70 °C

  14. Current Status of the Elevated Temperature Structure Design Codes for VHTR

    International Nuclear Information System (INIS)

    Kim, Jong-Bum; Kim, Seok-Hoon; Park, Keun-Bae; Lee, Won-Jae

    2006-01-01

    An elevated temperature structure design and analysis is one of the key issues in the VHTR (Very High Temperature Reactor) project to achieve an economic production of hydrogen which will be an essential energy source for the near future. Since the operating temperature of a VHTR is above 850 .deg. C, the existing code and standards are insufficient for a high temperature structure design. Thus the issues concerning a material selection and behaviors are being studied for the main structural components of a VHTR in leading countries such as US, France, UK, and Japan. In this study, the current status of the ASME code, French RCC-MR, UK R5, and Japanese code were investigated and the necessary R and D items were discussed

  15. The design of multi temperature and humidity monitoring system for incubator

    Science.gov (United States)

    Yu, Junyu; Xu, Peng; Peng, Zitao; Qiang, Haonan; Shen, Xiaoyan

    2017-01-01

    Currently, there is only one monitor of the temperature and humidity in an incubator, which may cause inaccurate or unreliable data, and even endanger the life safety of the baby. In order to solve this problem,we designed a multi-point temperature and humidity monitoring system for incubators. The system uses the STC12C5A60S2 microcontrollers as the sender core chip which is connected to four AM2321 temperature and humidity sensors. We select STM32F103ZET6 core development board as the receiving end,cooperating with Zigbee wireless transmitting and receiving module to realize data acquisition and transmission. This design can realize remote real-time observation data on the computer by communicating with PC via Ethernet. Prototype tests show that the system can effectively collect and display the information of temperature and humidity of multiple incubators at the same time and there are four monitors in each incubator.

  16. Steric Hindrance as a Basis for Structure-Based Design of Selective Inhibitors of Protein-Tyrosine Phosphatases

    DEFF Research Database (Denmark)

    Iversen, L. F.; Andersen, H. S.; Møller, K. B.

    2001-01-01

    Utilizing structure-based design, we have previously demonstrated that it is possible to obtain selective inhibitors of protein-tyrosine phosphatase 1B (PTP1B). A basic nitrogen was introduced into a general PTP inhibitor to form a salt bridge to Asp48 in PTP1B and simultaneously cause repulsion...... in PTPs containing an asparagine in the equivalent position [Iversen, L. F., et al. (2000) J. Biol. Chem. 275, 10300−10307]. Further, we have recently demonstrated that Gly259 in PTP1B forms the bottom of a gateway that allows easy access to the active site for a broad range of substrates, while bulky...... in accessibility to the active site among various PTPs. We show that a general, low-molecular weight PTP inhibitor can be developed into a highly selective inhibitor for PTP1B and TC-PTP by introducing a substituent, which is designed to address the region around residues 258 and 259. Detailed enzyme kinetic...

  17. Design, Experiments and Simulation of Voltage Transformers on the Basis of a Differential Input D-dot Sensor

    Directory of Open Access Journals (Sweden)

    Jingang Wang

    2014-07-01

    Full Text Available Currently available traditional electromagnetic voltage sensors fail to meet the measurement requirements of the smart grid, because of low accuracy in the static and dynamic ranges and the occurrence of ferromagnetic resonance attributed to overvoltage and output short circuit. This work develops a new non-contact high-bandwidth voltage measurement system for power equipment. This system aims at the miniaturization and non-contact measurement of the smart grid. After traditional D-dot voltage probe analysis, an improved method is proposed. For the sensor to work in a self-integrating pattern, the differential input pattern is adopted for circuit design, and grounding is removed. To prove the structure design, circuit component parameters, and insulation characteristics, Ansoft Maxwell software is used for the simulation. Moreover, the new probe was tested on a 10 kV high-voltage test platform for steady-state error and transient behavior. Experimental results ascertain that the root mean square values of measured voltage are precise and that the phase error is small. The D-dot voltage sensor not only meets the requirement of high accuracy but also exhibits satisfactory transient response. This sensor can meet the intelligence, miniaturization, and convenience requirements of the smart grid.

  18. Design, experiments and simulation of voltage transformers on the basis of a differential input D-dot sensor.

    Science.gov (United States)

    Wang, Jingang; Gao, Can; Yang, Jie

    2014-07-17

    Currently available traditional electromagnetic voltage sensors fail to meet the measurement requirements of the smart grid, because of low accuracy in the static and dynamic ranges and the occurrence of ferromagnetic resonance attributed to overvoltage and output short circuit. This work develops a new non-contact high-bandwidth voltage measurement system for power equipment. This system aims at the miniaturization and non-contact measurement of the smart grid. After traditional D-dot voltage probe analysis, an improved method is proposed. For the sensor to work in a self-integrating pattern, the differential input pattern is adopted for circuit design, and grounding is removed. To prove the structure design, circuit component parameters, and insulation characteristics, Ansoft Maxwell software is used for the simulation. Moreover, the new probe was tested on a 10 kV high-voltage test platform for steady-state error and transient behavior. Experimental results ascertain that the root mean square values of measured voltage are precise and that the phase error is small. The D-dot voltage sensor not only meets the requirement of high accuracy but also exhibits satisfactory transient response. This sensor can meet the intelligence, miniaturization, and convenience requirements of the smart grid.

  19. Applying Human-performance Models to Designing and Evaluating Nuclear Power Plants: Review Guidance and Technical Basis

    Energy Technology Data Exchange (ETDEWEB)

    O' Hara, J.M.

    2009-11-30

    Human performance models (HPMs) are simulations of human behavior with which we can predict human performance. Designers use them to support their human factors engineering (HFE) programs for a wide range of complex systems, including commercial nuclear power plants. Applicants to U.S. Nuclear Regulatory Commission (NRC) can use HPMs for design certifications, operating licenses, and license amendments. In the context of nuclear-plant safety, it is important to assure that HPMs are verified and validated, and their usage is consistent with their intended purpose. Using HPMs improperly may generate misleading or incorrect information, entailing safety concerns. The objective of this research was to develop guidance to support the NRC staff's reviews of an applicant's use of HPMs in an HFE program. The guidance is divided into three topical areas: (1) HPM Verification, (2) HPM Validation, and (3) User Interface Verification. Following this guidance will help ensure the benefits of HPMs are achieved in a technically sound, defensible manner. During the course of developing this guidance, I identified several issues that could not be addressed; they also are discussed.

  20. Applying Human-performance Models to Designing and Evaluating Nuclear Power Plants: Review Guidance and Technical Basis

    International Nuclear Information System (INIS)

    O'Hara, J.M.

    2009-01-01

    Human performance models (HPMs) are simulations of human behavior with which we can predict human performance. Designers use them to support their human factors engineering (HFE) programs for a wide range of complex systems, including commercial nuclear power plants. Applicants to U.S. Nuclear Regulatory Commission (NRC) can use HPMs for design certifications, operating licenses, and license amendments. In the context of nuclear-plant safety, it is important to assure that HPMs are verified and validated, and their usage is consistent with their intended purpose. Using HPMs improperly may generate misleading or incorrect information, entailing safety concerns. The objective of this research was to develop guidance to support the NRC staff's reviews of an applicant's use of HPMs in an HFE program. The guidance is divided into three topical areas: (1) HPM Verification, (2) HPM Validation, and (3) User Interface Verification. Following this guidance will help ensure the benefits of HPMs are achieved in a technically sound, defensible manner. During the course of developing this guidance, I identified several issues that could not be addressed; they also are discussed.

  1. Study of the Army Helicopter Design Hover Criterion Using Temperature and Pressure Altitude

    Science.gov (United States)

    2017-09-01

    DOCUMENTS, DESTROY BY ANY METHOD THAT WILL PREVENT DISCLOSURE OF CONTENTS OR RECONSTRUCTION OF THE DOCUMENT. DISCLAIMER THE FINDINGS IN THIS...ambient temperature design point [Lavallee and Sing, 1965]. However, this recommendation did not account for diurnal temperature variation. In 1975...altitude requirement to 6,000 feet while maintaining the 500 feet per minute VROC with 5 percent power margin capability to account for realistic

  2. Design rule for fatigue of welded joints in elevated-temperature nuclear components

    International Nuclear Information System (INIS)

    O'Connor, D.G.; Corum, J.M.

    1986-01-01

    Elevated-temperature weldment fatigue failures have occurred in several operating liquid-metal reactor plants. Yet, ASME Code Case N-47, which governs the design of such plants in the United States, does not currently address the Code Subgroup on Elevated Temperature Design recently proposed a fatigue strength reduction factor for austenitic and ferritic steel weldments. The factor is based on a variety of weld metal and weldment fatigue data generated in the United States, Europe, and Japan. This paper describes the factor and its bases, and it presents the results of confirmatory fatigue tests conducted at Oak Ridge National Laboratory on 316 stainless steel tubes with axial and circumferential welds of 16-8-2 filler metal. These test results confirm the suitability of the design factor, and they support the premise that the metallurgical notch effect produced by yield strength variations across a weldment is largely responsible for the observed elevated-temperature fatigue strength reduction

  3. Including severe accidents in the design basis of nuclear power plants: An organizational factors perspective after the Fukushima accident

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.; Frutuoso e Melo, P.F.

    2015-01-01

    Highlights: • The Fukushima accident was man-made and not caused by natural phenomena. • Vulnerabilities were known by regulator and licensee but measures were not taken. • There was lack of independence and transparency of the regulatory body. • Laws and regulations have not been updated to international standards. • Organizational failures have played an important role in the Fukushima accident. - Abstract: The Fukushima accident was clearly an accident made by humans and not caused by natural phenomena as was initially thought. Vulnerabilities were known by both regulators and operator but they postponed measures. The emergency plan was not effective in protecting the public, because the involved parties were not sufficiently prepared to make the right decisions. The shortcomings and faults mentioned above resulted from the lack of independence and transparency of the regulatory body. Even laws and regulations, and technical standards, have not been upgraded to international standards. Regulators have not defined requirements and left for the operator to decide what would be more appropriate. In this aspect, there was clearly a lack of independence between these bodies and operator’s lobby power. The above situation raised the question of urgent updating of institutions, in particular those responsible for nuclear safety. The above evidences show that several nuclear safety principles were not followed. This paper intends to highlight some existing safety criteria that were developed from the operational experience of the severe accidents that occurred at TMI and Chernobyl that should be incorporated in the design of new nuclear power plants and to provide appropriate design changes (backfittings) for reactors that belong to the previous generation prior to the occurrence of these accidents, through the study of design vulnerabilities. Furthermore, the main criteria that define an effective regulatory agency are also discussed. Although these

  4. Basis and Statistical Design of the Passive HIV-1 Antibody Mediated Prevention (AMP) Test-of-Concept Efficacy Trials.

    Science.gov (United States)

    Gilbert, Peter B; Juraska, Michal; deCamp, Allan C; Karuna, Shelly; Edupuganti, Srilatha; Mgodi, Nyaradzo; Donnell, Deborah J; Bentley, Carter; Sista, Nirupama; Andrew, Philip; Isaacs, Abby; Huang, Yunda; Zhang, Lily; Capparelli, Edmund; Kochar, Nidhi; Wang, Jing; Eshleman, Susan H; Mayer, Kenneth H; Magaret, Craig A; Hural, John; Kublin, James G; Gray, Glenda; Montefiori, David C; Gomez, Margarita M; Burns, David N; McElrath, Julie; Ledgerwood, Julie; Graham, Barney S; Mascola, John R; Cohen, Myron; Corey, Lawrence

    2017-01-01

    Anti-HIV-1 broadly neutralizing antibodies (bnAbs) have been developed as potential agents for prevention of HIV-1 infection. The HIV Vaccine Trials Network and the HIV Prevention Trials Network are conducting the Antibody Mediated Prevention (AMP) trials to assess whether, and how, intravenous infusion of the anti-CD4 binding site bnAb, VRC01, prevents HIV-1 infection. These are the first test-of-concept studies to assess HIV-1 bnAb prevention efficacy in humans. The AMP trials are two parallel phase 2b HIV-1 prevention efficacy trials conducted in two cohorts: 2700 HIV-uninfected men and transgender persons who have sex with men in the United States, Peru, Brazil, and Switzerland; and 1500 HIV-uninfected sexually active women in seven countries in sub-Saharan Africa. Participants are randomized 1:1:1 to receive an intravenous infusion of 10 mg/kg VRC01, 30 mg/kg VRC01, or a control preparation every 8 weeks for a total of 10 infusions. Each trial is designed (1) to assess overall prevention efficacy (PE) pooled over the two VRC01 dose groups vs. control and (2) to assess VRC01 dose and laboratory markers as correlates of protection (CoPs) against overall and genotype- and phenotype-specific infection. Each AMP trial is designed to have 90% power to detect PE > 0% if PE is ≥ 60%. The AMP trials are also designed to identify VRC01 properties (i.e., concentration and effector functions) that correlate with protection and to provide insight into mechanistic CoPs. CoPs are assessed using data from breakthrough HIV-1 infections, including genetic sequences and sensitivities to VRC01-mediated neutralization and Fc effector functions. The AMP trials test whether VRC01 can prevent HIV-1 infection in two study populations. If affirmative, they will provide information for estimating the optimal dosage of VRC01 (or subsequent derivatives) and identify threshold levels of neutralization and Fc effector functions associated with high-level protection, setting a benchmark

  5. Produced Water Treatment Using the Switchable Polarity Solvent Forward Osmosis (SPS FO) Desalination Process: Preliminary Engineering Design Basis

    Energy Technology Data Exchange (ETDEWEB)

    Wendt, Daniel; Adhikari, Birendra; Orme, Christopher; Wilson, Aaron

    2016-05-01

    Switchable Polarity Solvent Forward Osmosis (SPS FO) is a semi-permeable membrane-based water treatment technology. INL is currently advancing SPS FO technology such that a prototype unit can be designed and demonstrated for the purification of produced water from oil and gas production operations. The SPS FO prototype unit will used the thermal energy in the produced water as a source of process heat, thereby reducing the external process energy demands. Treatment of the produced water stream will reduce the volume of saline wastewater requiring disposal via injection, an activity that is correlated with undesirable seismic events, as well as generate a purified product water stream with potential beneficial uses. This paper summarizes experimental data that has been collected in support of the SPS FO scale-up effort, and describes how this data will be used in the sizing of SPS FO process equipment. An estimate of produced water treatment costs using the SPS FO process is also provided.

  6. Design and Application of a High-Temperature Linear Ion Trap Reactor

    Science.gov (United States)

    Jiang, Li-Xue; Liu, Qing-Yu; Li, Xiao-Na; He, Sheng-Gui

    2018-01-01

    A high-temperature linear ion trap reactor with hexapole design was homemade to study ion-molecule reactions at variable temperatures. The highest temperature for the trapped ions is up to 773 K, which is much higher than those in available reports. The reaction between V2O6 - cluster anions and CO at different temperatures was investigated to evaluate the performance of this reactor. The apparent activation energy was determined to be 0.10 ± 0.02 eV, which is consistent with the barrier of 0.12 eV calculated by density functional theory. This indicates that the current experimental apparatus is prospective to study ion-molecule reactions at variable temperatures, and more kinetic details can be obtained to have a better understanding of chemical reactions that have overall barriers. [Figure not available: see fulltext.

  7. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    2004-07-29

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

  8. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    International Nuclear Information System (INIS)

    Ingersoll, D.T.

    2004-01-01

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

  9. Design of cross-sensitive temperature and strain sensor based on sampled fiber grating

    Directory of Open Access Journals (Sweden)

    Zhang Xiaohang

    2017-02-01

    Full Text Available In this paper,a cross-sensitive temperature and strain sensor based on sampled fiber grating is designed.Its temperature measurement range is -50-200℃,and the strain measurement rangeis 0-2 000 με.The characteristics of the sensor are obtained using simulation method.Utilizing SPSS software,we found the dual-parameter matrix equations of measurement of temperature and strain,and calibrated the four sensing coefficients of the matrix equations.

  10. Design of a termination for a high temperature superconduction power cable

    DEFF Research Database (Denmark)

    Rasmussen, Carsten; Kühle (fratrådt), Anders Van Der Aa; Tønnesen, Ole

    1999-01-01

    ). This assembly is electrically insulated with an extruded polymer dielectric kept at room temperature. Cooling is provided by a flow of liquid nitrogen inside the former. The purpose of the termination is to connect the superconducting cable conductor at cryogenic temperature to the existing power grid at room...... temperatures, the transfer of liquid nitrogen over a high voltage drop and that of providing a well defined atmosphere inside the termination and around the cable conductor. Designs based on calculations and experiments will be presented. The solutions are optimized with respect to a low heat in-leak....

  11. Design, fabrication and characterisation of a microfluidic time-temperature indicator

    Science.gov (United States)

    Schmitt, P.; Wedrich, K.; Müller, L.; Mehner, H.; Hoffmann, M.

    2017-11-01

    This paper describes a concept for a passive microfluidic time-temperature indicator (TTI) intended for intelligent food packaging. A microfluidic system is presented that makes use of the temperature-dependent flow of suitable food ingredients in a microcapillary. Based on the creeping distance inside the capillary, the time-temperature integral can be determined. A demonstrator of the microsystem has been designed, fabricated and characterised using liquid sugar alcohols as indicator fluids. To enable a first wireless read-out of the passive TTI, the sensor was read out using a commercial RFID equipment, and capacitive measurements have been carried out.

  12. On subcooler design for integrated two-temperature supermarket refrigeration system

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Liang [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Zhang, Chun-Lu [College of Mechanical Engineering, Tongji University, No. 4800, Cao An Highway, Shanghai 201804 (China)

    2011-01-15

    The energy saving opportunity of supermarket refrigeration systems using subcooler between the medium-temperature (MT) refrigeration system and the low-temperature (LT) refrigeration system has been identified in the previous work. This paper presents a model-based comprehensive analysis on the subcooler design. The optimal subcooling control is discussed as well. With optimal subcooler size and subcooling control, the maximum energy savings of integrated two-temperature supermarket refrigeration system using R404A or R134a as working fluid can achieve 27% or 20%, respectively. The load ratio of MT to LT system and the operating conditions have considerable impact on the energy savings. (author)

  13. Temperature modeling for analysis and design of the sintering furnance in HTR fuel type of ball

    International Nuclear Information System (INIS)

    Saragi, Elfrida; Setiadji, Moch

    2013-01-01

    One of the factors that determine the safety of the operation of the sintering furnace fuel HTR ball is the temperature distribution in the ceramic tube furnace. The temperature distribution must be determined at design stage. The tube has a temperature of 1600 °C at one end and about 40 °C at the other end. The outside of the tube was cooled by air through natural convection. The tube is a furnace ceramic tube which its geometry are 0.08, 0.09 and 0.5 m correspondingly for the inner tube diameter, outer tube diameter and tube length. The temperature distribution of the tube is determined by the natural convection coefficient (NCF), which is difficult to be calculated manually. The determination of NCF includes the Grasshoff, Prandtl, and Nusselt numbers which is a function of the temperature difference between the surrounding air with the ceramic tube. If the temperature vary along the tube, the complexity of the calculations increases. Thus the proposed modeling was performed to determine the temperature distribution along the tube and heat transfer coefficient using a self-developed software which permit the design process easier

  14. Conceptual design of a municipal energy and environmental system as an efficient basis for advanced energy planning

    International Nuclear Information System (INIS)

    Kostevšek, Anja; Petek, Janez; Čuček, Lidija; Pivec, Aleksandra

    2013-01-01

    Effectively implementing various energy and environmental policies contributes to the acceleration of energy performance, a reduction in negative environmental impacts, and increased deployment of renewable resources. The MEEMS (municipal energy and environmental management system) performs the almost inconceivable role of accomplishing prerequisite targets at the national level and, consequently, the European and World levels also. Therefore, a proper infrastructure for MEEMS needs to be effectively applied in order to implement policy initiatives. A novel organisational framework of MEEMS is proposed and is constituted upon three pillars: integration of the municipal metabolism approach, the KBS (knowledge-based system), and the MTIS (municipal technology innovation system). By properly addressing the dynamics of the MEES (municipal energy and environmental system), and the new conceptual organisation of MEEMS, a need for the inclusion of innovative elements can be defined regarding support mechanisms. Integration of the end-user approach defines the fundamental orientation of modern MEEMS. This new concept paves a pathway towards an intelligent energy and environmental system. This paper describes an implementation of the new conceptual design of MEEMS within the urban municipality energy system of Ptuj, Slovenia. - Highlights: • Combination of energy and environmental management system (MEEMS) within municipalities is proposed. • Novel MEEMS structure to accomplish more effective functioning of the system. • Accelerated energy policy implementation process is enabled with proposed MEEMS. • Key ingredients to transform present energy systems to more innovative ones are discussed

  15. Design basis knowledge management for the newcomer countries. Relying on the owner/operator as a knowledgeable customer

    International Nuclear Information System (INIS)

    Lepouze, Benoît

    2013-01-01

    Becoming a knowledgeable customer is the first step to manage knowledge: • Vendors, consulting firms, TSOs can assist the future operator but it will remain the sole owner of the decisions; • Future owner operator has to become a knowledgeable customer: • Know what to ask for, • Know how to ask for it, • Know how to check if it got what it asked. • Where should knowledge management belong (management?, HRD?, procurement?) and is it important? What it means for DBKM ? (example): • The owner/operator (the licensee) is responsible in front of the safety agency: it should answer its questions at every stage of the program; • It will often turn back to its vendor/suppliers especially for detailed design questions; • But that means it has to know what to ask for and to check the result before talking to the regulator; • That also means it has to make sure knowledge is managed throughout the life of the program

  16. An assessment of the radiological consequences of the Greek Research Reactor's design basis accident with the use of low enriched uranium

    International Nuclear Information System (INIS)

    Kollas, J.G.

    1985-09-01

    An analysis of the radiological consequences of the design basis accident in the low enriched uranium fueled 5 MW Greek Research Reactor is presented. For the source term thirty-five isotopes are taken into consideration and conservative figures of fission product release are adopted. To estimate the reactor's consequences for Athens population a CRAC2 consequence model version is used. The results indicate that limiting dose and effects are respectively the thyroid dose and the thyroid effects induced in the 3,081,000 inhabitants of Athens region. (author)

  17. Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

    International Nuclear Information System (INIS)

    Mouri, Tomoaki; Kunitomi, Kazuhiko

    2008-01-01

    A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

  18. Key technological issues in LMFBR high-temperature structural design - the US perspective

    International Nuclear Information System (INIS)

    Corum, J.M.

    1984-01-01

    The purpose of this paper is: (1) to review the key technological issues in LMFBR high-temperature structural design, particularly as they relate to cost reduction; and (2) to provide an overview of activities sponsored by the US Department of Energy to resolve the issues and to establish stable, standardized, and defensible structural design methods and criteria. Specific areas of discussion include: weldments, structural validation tests, simplified design analysis procedures, design procedures for piping, validation of the methodology for notch-like geometries, improved life assessment procedures, thermal striping, extension of the methodology to new materials, and ASME high-temperature Code reform needs. The perceived problems and needs in each area are discussed, and the current status of related US activities is given

  19. Nonlinear structural analysis methods and their application to elevated temperature design: A US perspective

    International Nuclear Information System (INIS)

    Dhalla, A.K.

    1989-01-01

    Technological advances over the last two decades have been assimilated into the routine design of Liquid Metal Reactor (LMR) structural components operating at elevated temperatures. The mature elevated temperature design technology is based upon: (a) an extensive material data base, (b) recent advances in nonlinear computational methods, and (c) conservative design criteria based upon past successful and reliable operating experiences with petrochemical and nonnuclear power plants. This survey paper provides a US perspective on the role of nonlinear analysis methods used in the design of LMR plants. The simplified and detailed nonlinear analysis methods and the level of computational effort required to qualify structural components for safe and reliable long-term operation are discussed. The paper also illustrates how a detailed nonlinear analysis can be used to resolve technical licensing issues, to understand complex nonlinear structural behavior, to identify predominant failure modes, and to guide future experimental programs

  20. Design and analysis of push pipe joint under internal pressure and temperature loading

    International Nuclear Information System (INIS)

    Abid, M.; Alam, K.

    2005-01-01

    Pipe joints flanged or welded are commonly used in industry for different applications ranging from sewerage to the high pressure and temperature applications. However, with the rapidly changing technological trends, for optimized space such as for heat exchanger applications, pipe joint design needs special consideration, especially for the internal pipe where no flanged/bolted joint due to space constraint can be used. In addition, where joint opening/closing is the requirement for maintenance or other functional purposes, it becomes inevitable to use some special design. In this paper, a push joint proposed is designed, analyzed, optimized and tested for safe stress and operating conditions. An experimental test rig is designed and tests are performed for internal pressure and temperature separately and joint's behaviour is analyzed in detail for any leaks. FEA results are compared and verified with the mathematical results. Based on the experimental observations, the joint is safe as no leaks are observed. (author)

  1. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Haddam Neck Nuclear Power Plant

    International Nuclear Information System (INIS)

    Laudenbach, D.H.

    1979-03-01

    The technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Haddam Neck Nuclear Power Plant is presented. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Support Program being conducted for the U.S. Nuclear Regulatory Commission by Lawrence Livermore Laboratory

  2. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Point Beach Nuclear Power Plant, Units 1 and 2

    International Nuclear Information System (INIS)

    Laudenbach, D.H.

    1979-03-01

    The technical evaluation is presented for the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Point Beach nuclear power plant, Units 1 and 2. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Support Program being conducted for the U.S. Nuclear Regulatory Commission by Lawrence Livermore Laboratory

  3. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Point Beach Nuclear Power Plant, Units 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Laudenbach, D.H.

    1979-03-01

    The technical evaluation is presented for the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Point Beach nuclear power plant, Units 1 and 2. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Support Program being conducted for the U.S. Nuclear Regulatory Commission by Lawrence Livermore Laboratory.

  4. Theoretical basis, principles of design, and experimental study of the prototype of perfect AFCS transmitting signals without coding

    Science.gov (United States)

    Platonov, A.; Zaitsev, Ie.; Opalski, L. J.

    2017-08-01

    The paper presents an overview of design methodology and results of experiments with a Prototype of highly efficient optimal adaptive feedback communication systems (AFCS), transmitting low frequency analog signals without coding. The paper emphasizes the role of the forward transmitter saturation as the factor that blocked implementation of theoretical results of pioneer (1960-1970s) and later research on FCS. Deepened analysis of the role of statistical fitting condition in adequate formulation and solution of AFCS optimization task is given. Solution of the task - optimal transmission/reception algorithms is presented in the form useful for elaboration of the hardware/software Prototype. A notable particularity of the Prototype is absence of the encoding/decoding units, whose functions are realized by the adaptive pulse amplitude modulator (PAM) of the forward transmitter (FT) and estimating/controlling algorithm in the receiver of base station (BS). Experiments confirm that the Prototype transmits signals from FT to BS "perfectly": with the bit rate equal to the capacity of the system, and with limit energy [J/bit] and spectral [bps/Hz] efficiency. Another, not less important and confirmed experimentally, particularity of AFCS is its capability to adjust parameters of FT and BS to the characteristics of scenario of application and maintain the ideal regime of transmission including spectralenergy efficiency. AFCS adjustment can be made using BS estimates of mean square error (MSE). The concluding part of the paper contains discussion of the presented results, stressing capability of AFCS to solve problems appearing in development of dense wireless networks.

  5. Extreme temperature robust optical sensor designs and fault-tolerant signal processing

    Science.gov (United States)

    Riza, Nabeel Agha [Oviedo, FL; Perez, Frank [Tujunga, CA

    2012-01-17

    Silicon Carbide (SiC) probe designs for extreme temperature and pressure sensing uses a single crystal SiC optical chip encased in a sintered SiC material probe. The SiC chip may be protected for high temperature only use or exposed for both temperature and pressure sensing. Hybrid signal processing techniques allow fault-tolerant extreme temperature sensing. Wavelength peak-to-peak (or null-to-null) collective spectrum spread measurement to detect wavelength peak/null shift measurement forms a coarse-fine temperature measurement using broadband spectrum monitoring. The SiC probe frontend acts as a stable emissivity Black-body radiator and monitoring the shift in radiation spectrum enables a pyrometer. This application combines all-SiC pyrometry with thick SiC etalon laser interferometry within a free-spectral range to form a coarse-fine temperature measurement sensor. RF notch filtering techniques improve the sensitivity of the temperature measurement where fine spectral shift or spectrum measurements are needed to deduce temperature.

  6. Design study of the experimental multi-purpose high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Tsunoda, Ryokichi

    1981-01-01

    In this paper, the design study carried out since 1973 is outlined. The basic conceptual design was performed in fiscal 1973. In this design, concept was established on the total system of the experimental high temperature gas-cooled reactor including heat-utilizing system. The first conceptual design was carried out in fiscal 1974. The range of design was limited to the experimental reactor and its direct heat-removing system. The part 2 of the first conceptual design was performed in fiscal 1975, and the system design concerning the plant characteristics was made. The part 1 of the adjustment design was carried out in fiscal 1976, and the subject was the adjustment design of plant systems. The part 2 was performed in fiscal 1977, and the characteristics of plant control system were analyzed. In fiscal 1978, the analysis of flow characteristics in the core was made. The integrated system design was carried out in fiscal 1979, and the design of the total plant system except heat-utilizing system was started again. The part 1 of the detailed design was performed in fiscal 1980, and in addition, the possibility of increasing power output was examined. The construction cost of the experimental reactor plant estimated in 1979 was far higher than that in 1973. (Kako, I.)

  7. Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

    2006-12-15

    The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA.

  8. Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

    International Nuclear Information System (INIS)

    Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

    2006-12-01

    The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA

  9. Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyeong-Yeon, E-mail: hylee@kaeri.re.kr

    2016-11-15

    Highlights: • Comparison of elevated temperature design (ETD) codes was made. • Material properties and evaluation procedures were compared. • Two heat-resistant materials of Grade 91 steel and austenitic stainless steel 316 are the target materials in the present study. • Application of the ETD codes to Generation IV reactor components and a comparison of the conservatism was conducted. - Abstract: The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.

  10. Discussion on Design Transients of Pebble-bed High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Wang Yan; Li Fu; Zheng Yanhua

    2014-01-01

    In order to assure high quality for the components and their supports in the reactor coolant system, etc., some thermal-hydraulic transient conditions will be selected and researched for equipment design evaluation to satisfy the requirements ASME code, which are based on the conservative estimates of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in the plant. In the mature design on pressurized water reactor, five conditions are considered. For the developing advanced pebble-bed high temperature gas-cooled reactor(HTGR), its design and operation has much difference with other reactors, so the transients of the pebble-bed high temperature gas-cooled reactor have distinctive characteristics. In this paper, the possible design transients of the pebble-bed HTGR will be discussed, and the frequency of design transients for equipment fatigue analysis and stress analysis due to cyclic stresses is also studied. The results will provide support for the design and construct of the pebble-bed HTGR. (author)

  11. Research design and improvement of high temperature high pressure solenoid valve

    International Nuclear Information System (INIS)

    Luo Yongtang

    1987-12-01

    A process for development of the pilot type high temperature high pressure solenoid valve used in a PWR power plant is described. The whole development process might be divided into two phases: research design and improvement. In the former phase the questions had chiefly been approached in the following several aspects: the principle construction design, the determination of values for the constructionally key elements, the valve seal design and the solenoid actuator design, and made such valve's successful design in the main. In the latter phase an improvement had been made upon such valve against the problems during the testing use of the valve for a period of time, i.e. the unsatisfactory leak tightness, and achieved satisfactory results. The consummate success in this development not only has met the needs of the engineering project, but also made us obtain a valuable experience useful to design the similar valves

  12. Design of Annular Linear Induction Pump for High Temperature Liquid Lead Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Jae Sik; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    EM(Electro Magnetic) Pump is divided into two parts, which consisted of the primary one with electromagnetic core and exciting coils, and secondary one with liquid lead flow. The main geometrical variables of the pump included core length, inner diameter and flow gap while the electromagnetic ones covered pole pitch, turns of coil, number of pole pairs, input current and input frequency. The characteristics of design variables are analyzed by electrical equivalent circuit method taking into account hydraulic head loss in the narrow annular channel of the ALIP. The design program, which was composed by using MATLAB language, was developed to draw pump design variables according to input requirements of the flow rate, developing pressure and operation temperature from the analyses. The analysis on the design of ALIP for high temperature liquid lead transportation was carried for the produce of ALIP designing program based on MATLAB. By the using of ALIP designing program, we don't have to bother about geometrical relationship between each component during detail designing process because code calculate automatically. And prediction of outputs about designing pump can be done easily before manufacturing. By running the code, we also observe and analysis change of outputs caused by changing of pump factors. It will be helpful for the research about optimization of pump outputs.

  13. Creative design-by-analysis solutions applied to high-temperature components

    International Nuclear Information System (INIS)

    Dhalla, A.K.

    1993-01-01

    Elevated temperature design has evolved over the last two decades from design-by-formula philosophy of the ASME Boiler and Pressure Vessel Code, Sections I and VIII (Division 1), to the design-by-analysis philosophy of Section III, Code Case N-47. The benefits of design-by-analysis procedures, which were developed under a US-DOE-sponsored high-temperature structural design (HTSD) program, are illustrated in the paper through five design examples taken from two U.S. liquid metal reactor (LMR) plants. Emphasis in the paper is placed upon the use of a detailed, nonlinear finite element analysis method to understand the structural response and to suggest design optimization so as to comply with Code Case N-47 criteria. A detailed analysis is cost-effective, if selectively used, to qualify an LMR component for service when long-lead-time structural forgings, procured based upon simplified preliminary analysis, do not meet the design criteria, or the operational loads are increased after the components have been fabricated. In the future, the overall costs of a detailed analysis will be reduced even further with the availability of finite element software used on workstations or PCs

  14. New design of a variable-temperature ultrahigh vacuum scanning tunneling microscope

    NARCIS (Netherlands)

    Mugele, Friedrich Gunther; Rettenberger, A.; Boneberg, J.; Leiderer, P.

    1998-01-01

    We present the design of a variable-temperature ultrahigh vacuum (UHV) scanning tunneling microscope which can be operated between 20 and 400 K. The microscope is mounted directly onto the heat exchanger of a He continuous flow cryostat without vibration isolation inside the UHV chamber. The coarse

  15. Design activity of IHI on the experimental multipurpose high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    1978-01-01

    With conspicuous interest and attention paid by iron and steel manufacturing industries, the development of the multipurpose high temperature gas-cooled reactor, namely the process heat reactor has been energetically discussed in Japan. The experimental multipurpose high temperature gas-cooled reactor, planned by JAERI (the Japan Atomic Energy Research Institute), is now at the end of the adjustment design stage and about to enter the system synthesizing design stage. The design of the JAERI reactor as a pilot plant for process heat reactors that make possible the direct use of the heat, produced in the reactor, for other industrial uses was started in 1969, and has undergone several revisions up to now. The criticality of the JAERI reactor is expected to be realized before 1985 according to the presently published program. IHI has engaged in the developing work of HTGR (high temperature gas-cooled reactor) including VHTR (very high temperature gas-cooled reactor) for over seven years, producing several achievements. IHI has also participated in the JAERI project since 1973 with some other companies concerned in this field. The design activity of IHI in the development of the JAERI reactor is briefly presented in this paper. (auth.)

  16. A novel design for a cheap high temperature solar collector: The rotating solar boiler

    NARCIS (Netherlands)

    Luijtelaer, van J.P.H.; Kroon, M.C.

    2009-01-01

    In this work a novel type of high temperature solar collector is designed: the rotating solar boiler. This rotating solar boiler consists of two concentric tubes. The inner tube, called absorber, absorbs sunlight and boils water. The outer transparent tube, called cover, is filled with air. The

  17. Design and modelling of a novel compact power cycle for low temperature heat sources

    DEFF Research Database (Denmark)

    Wronski, Jorrit; Skovrup, Morten Juel; Elmegaard, Brian

    2012-01-01

    Power cycles for the efficient use of low temperature heat sources experience increasing attention. This paper describes an alternative cycle design that offers potential advantages in terms of heat source exploitation. A concept for a reciprocating expander is presented that performs both, work ...

  18. Design of project management system for 10 MW high temperature gas-cooled test reactor

    International Nuclear Information System (INIS)

    Zhu Yan; Xu Yuanhui

    1998-01-01

    A framework of project management information system (MIS) for 10 MW high temperature gas-cooled test reactor is introduced. Based on it, the design of nuclear project management information system and project monitoring system (PMS) are given. Additionally, a new method of developing MIS and Decision Support System (DSS) has been tried

  19. Automatic performance estimation of conceptual temperature control system design for rapid development of real system

    International Nuclear Information System (INIS)

    Jang, Yu Jin

    2013-01-01

    This paper presents an automatic performance estimation scheme of conceptual temperature control system with multi-heater configuration prior to constructing the physical system for achieving rapid validation of the conceptual design. An appropriate low-order discrete-time model, which will be used in the controller design, is constructed after determining several basic factors including the geometric shape of controlled object and heaters, material properties, heater arrangement, etc. The proposed temperature controller, which adopts the multivariable GPC (generalized predictive control) scheme with scale factors, is then constructed automatically based on the above model. The performance of the conceptual temperature control system is evaluated by using a FEM (finite element method) simulation combined with the controller.

  20. The Composition and Temperature Effects on the Ultra High Strength Stainless Steel Design

    Science.gov (United States)

    Xu, W.; Del Castillo, P. E. J. Rivera Díaz; van der Zwaag, S.

    Alloy composition and heat treatment are of paramount importance to determining alloy properties. Their control is of great importance for new alloy design and industrial fabrication control. A base alloy utilizing MX carbide is designed through a theory guided computational approach coupling a genetic algorithm with optimization criteria based on thermodynamic, kinetic and mechanical principles. The combined effects of 11 alloying elements (Al, C, Co, Cr, Cu, Mo, Nb, Ni, Si, Ti and V) are investigated in terms of the composition optimization criteria: the martensite start (Ms) temperature, the suppression of undesirable phases, the Cr concentration in the matrix and the potency of the precipitation strengthening contribution. The results show the concentration sensitivities of each component and also point out new potential composition domains for further strength increase. The aging temperature effect is studied and the aging temperature industrially followed is recovered.