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Sample records for standard design basis

  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. 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)

  3. 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.

  4. 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

  5. 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.

  6. 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)

  7. 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.

  8. 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...

  9. 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

  10. 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...

  11. 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

  12. 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.)

  13. Design basis for resistance to shock and vibration of radioactive material packages greater than one ton in truck transport (draft standard for trial use and comment)

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    This standard specifies minimum design values for shock and vibration in highway transport, by truck or by tractor-trailer combination, for fuel and irradiation experiments when package weight exceeds one ton. Shock values correspond to normal transport over rough roads and to minor accidents such as backing into a loading dock. Vibration values correspond to normal transport; any large-amplitude vibration resulting from rough road conditions or a minor accident is treated as shock. This standard includes recommended methods of application to the design of packaging and tiedown systems

  14. Geometrical basis for the Standard Model

    Science.gov (United States)

    Potter, Franklin

    1994-02-01

    The robust character of the Standard Model is confirmed. Examination of its geometrical basis in three equivalent internal symmetry spaces-the unitary plane C 2, the quaternion space Q, and the real space R 4—as well as the real space R 3 uncovers mathematical properties that predict the physical properties of leptons and quarks. The finite rotational subgroups of the gauge group SU(2) L × U(1) Y generate exactly three lepton families and four quark families and reveal how quarks and leptons are related. Among the physical properties explained are the mass ratios of the six leptons and eight quarks, the origin of the left-handed preference by the weak interaction, the geometrical source of color symmetry, and the zero neutrino masses. The ( u, d) and ( c, s) quark families team together to satisfy the triangle anomaly cancellation with the electron family, while the other families pair one-to-one for cancellation. The spontaneously broken symmetry is discrete and needs no Higgs mechanism. Predictions include all massless neutrinos, the top quark at 160 GeV/ c 2, the b' quark at 80 GeV/ c 2, and the t' quark at 2600 GeV/ c 2.

  15. 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

  16. 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.)

  17. 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.)

  18. 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

  19. 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...

  20. The biological basis of plutonium safety standards

    International Nuclear Information System (INIS)

    Mole, R.H.

    1976-01-01

    Since no radiation injury or cancer in man can, as yet, be directly attributed to Pu, all safety standards for Pu must be determined by reference to other safety standards, development of which is discussed. A system of safety standards must be based on links with real damage, such as the requirement for 226 Ra in bone. The type of biological information required for making standards realistic is considered in relation to Pu and Ra in bone. Also considered are the possible effects of Pu in soft tissue such as bone marrow. Not only dose, but also the number of cells exposed to the dose are important biologically and cellular aspects are examined. Since there is no positive evidence of Pu toxicity relevant information on other α emitters must be examined. The observed effectiveness of Ra, daughters of 222 Ra and 232 Th in causing mutations and cancer, is surveyed. Reference is made to the necessity of improving the ICRP system, currently based on the critical organ concept, by recognising the need for summation of risks in other organs where exposure to Pu is concerned. Improved biological understanding particularly that of hereditary damage, in recent years leads to less pessimistic thinking on the effects of ionizing radiations. The immediate need appears to be for consistency in safety standards. (U.K.)

  1. 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)

  2. 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

  3. Standards of Multimedia Graphic Design in Education

    Science.gov (United States)

    Aldalalah, Osamah Ahmad; Ababneh, Ziad Waleed Mohamed

    2015-01-01

    This study aims to determine Standards of Multimedia Graphic Design in Education through the analysis of the theoretical basis and previous studies related to this subject. This study has identified the list of standards of Multimedia, Graphic Design, each of which has a set indicator through which the quality of Multimedia can be evaluated in…

  4. 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

  5. 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)

  6. Working group 1A - basis for the standard-safety

    International Nuclear Information System (INIS)

    Whipple, C.

    1993-01-01

    This paper presents a summary of the progress made by working group 1A (Basis for the Safety Standard) during the Electric Power Research Institute's EPRI Workshop on the technical basis of EPA HLW Disposal Criteria, March 1993. This group discussed the semantics of terms within the standard 40 CFR Part 191, the implementation of this standard, the advanced notice of rulemaking, the issue of emitting carbon-14 through a gaseous pathway, the strategy of dealing with standards for contamination of drinking water and groundwater, the 100,000 year time frame, and the analysis of specific comments. The specific comments dealt with the cost effectiveness of the standard, the dose histogram for populations and individuals, groundwater definition and the underlying technology driver for this standard

  7. [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.

  8. 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

  9. 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

  10. Basis and rational for standardization in radiation protection

    Energy Technology Data Exchange (ETDEWEB)

    Besar, Idris

    1985-08-01

    The historical background for the standardization in radiation protection with special reference to the dose limits recommended by the ICRP which include tolerance dose, maximum permissible dose, and the present recommendation based on the ICRP 26 are presented. The basis and rational for the establishment of these limits are discussed.

  11. Basis and rational for standardization in radiation protection

    International Nuclear Information System (INIS)

    Idris Besar

    1985-01-01

    The historical background for the standardization in radiation protection with special reference to the dose limits recommended by the ICRP which include tolerance dose, maximum permissible dose, and the present recommendation based on the ICRP 26 are presented. The basis and rational for the establishement of these limits are discussed

  12. 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.

  13. 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

  14. Civil Engineering & Design Standards Manual

    OpenAIRE

    Vänttinen, Eetu

    2014-01-01

    Civil Discipline Engineering department in Foster Wheeler Energia Oy takes care of the construction of foundation, steel frame, platforms, cladding/roofing, HVAC, elevator, hoist and central vacuum system of the boiler building. The goal of the thesis was to compile a design manual for the department to ease up the startup of the design of a new project and standardize the design. Main objective was to gather together all the existing guidelines, standards and directives regarding the des...

  15. 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)

  16. 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

  17. 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.

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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)

  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. 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

  6. 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

  7. Radiobiological basis for setting neutron radiation safety standards

    International Nuclear Information System (INIS)

    Straume, T.

    1985-01-01

    Present neutron standards, adopted more than 20 yr ago from a weak radiobiological data base, have been in doubt for a number of years and are currently under challenge. Moreover, recent dosimetric re-evaluations indicate that Hiroshima neutron doses may have been much lower than previously thought, suggesting that direct data for neutron-induced cancer in humans may in fact not be available. These recent developments make it urgent to determine the extent to which neutron cancer risk in man can be estimated from data that are available. Two approaches are proposed here that are anchored in particularly robust epidemiological and experimental data and appear most likely to provide reliable estimates of neutron cancer risk in man. The first approach uses gamma-ray dose-response relationships for human carcinogenesis, available from Nagasaki (Hiroshima data are also considered), together with highly characterized neutron and gamma-ray data for human cytogenetics. When tested against relevant experimental data, this approach either adequately predicts or somewhat overestimates neutron tumorigenesis (and mutagenesis) in animals. The second approach also uses the Nagasaki gamma-ray cancer data, but together with neutron RBEs from animal tumorigenesis studies. Both approaches give similar results and provide a basis for setting neutron radiation safety standards. They appear to be an improvement over previous approaches, including those that rely on highly uncertain maximum neutron RBEs and unnecessary extrapolations of gamma-ray data to very low doses. Results suggest that, at the presently accepted neutron dose limit of 0.5 rad/yr, the cancer mortality risk to radiation workers is not very different from accidental mortality risks to workers in various nonradiation occupations

  8. MAN-004 Design Standards Manual

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Timothy L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-07-01

    At Sandia National Laboratories in New Mexico (SNL/NM), the design, construction, operation, and maintenance of facilities is guided by industry standards, a graded approach, and the systematic analysis of life cycle benefits received for costs incurred. The design of the physical plant must ensure that the facilities are "fit for use," and provide conditions that effectively, efficiently, and safely support current and future mission needs. In addition, SNL/NM applies sustainable design principles, using an integrated whole-building design approach, from site planning to facility design, construction, and operation to ensure building resource efficiency and the health and productivity of occupants. The safety and health of the workforce and the public, any possible effects on the environment, and compliance with building codes take precedence over project issues, such as performance, cost, and schedule. These design standards generally apply to all disciplines on all SNL/NM projects. Architectural and engineering design must be both functional and cost-effective. Facility design must be tailored to fit its intended function, while emphasizing low-maintenance, energy-efficient, and energy-conscious design. Design facilities that can be maintained easily, with readily accessible equipment areas, low maintenance, and quality systems. To promote an orderly and efficient appearance, architectural features of new facilities must complement and enhance the existing architecture at the site. As an Architectural and Engineering (A/E) professional, you must advise the Project Manager when this approach is prohibitively expensive. You are encouraged to use professional judgment and ingenuity to produce a coordinated interdisciplinary design that is cost-effective, easily contractible or buildable, high-performing, aesthetically pleasing, and compliant with applicable building codes. Close coordination and development of civil, landscape, structural, architectural, fire

  9. Seismic design standardization of nuclear facilities

    International Nuclear Information System (INIS)

    Reddy, G.R.; Vaze, K.K.

    2011-01-01

    Full text: Structures, Systems and Components (SSCs) of Nuclear Facilities have to be designed for normal operating loads such as dead weight, pressure, temperature etc., and accidental loads such as earthquakes, floods, extreme, wind air craft impact, explosions etc. Man made accidents such as aircraft impact, explosions etc., some times may be considered as design basis event and some times taken care by providing administrative controls. This will not be possible in the case of natural events such as earthquakes, flooding, extreme winds etc. Among natural events earthquakes are considered as most devastating and need to be considered as design basis event. It is generally felt design of SSCs for earthquake loads is very time consuming and expensive. Conventional seismic design approaches demands for large number of supports for systems and components. This results in large space occupation and in turn creates difficulties for maintenance and in service inspection of systems and components. In addition, complete exercise of design need to be repeated for plants being located at different sites due to different seismic demands. However, advanced seismic response control methods will help to standardize the seismic design meeting the safety and economy. These methods adopt passive, semi active and active devices, and base isolators to control the seismic response. In nuclear industry, it is advisable to go for passive devices to control the seismic responses. Ideally speaking, these methods will make the designs made for normal loads can also satisfy the seismic demand without calling for change in material, geometry, layout etc. in the SSCs. This paper explain the basic ideas of seismic response control methods, demonstrate the effectiveness of control methods through case studies and eventually give the procedure to be adopted for seismic design standardization of nuclear facilities

  10. 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.

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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.

  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. 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

  2. Six Dimension Strategy As A Basis Of Banking Standard Contract

    Directory of Open Access Journals (Sweden)

    Wulanmas Frederik

    2012-01-01

    Full Text Available Indonesia banking based on Article 4 of Act No.10, 1998, aims at supporting the implementation of national development in order to improve equity, economic growth and national stability in the direction of improving people’s welfare. Therefore, to show how important is banking role in supporting the implementation of development, the 6 (six Strategic Dimensions as the foundation of Banking Standards Contract are: (1. Prudent Banking Supervision and Good Corporate Governance (GCG in banking activities, (2. Refunctionalization the principle of Contract Law in Banking Standards Contract, (3. Ethics Value in Business, (4. The Act No. 8, 1999 on Consumer Protection, (5. Enforcement of Human Rights Principles in banking activities, (6. Abuse of Circumstances implementations (Misbruik van Omstandigheden in banking Contract. Based on the 6 (six Strategic Dimension as the foundation of Banking Standard Contract, it will undoubtedly create justice, equity and assurance of the rights and obligations of the parties framed in the contractual and law bonds.

  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. 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)

  5. 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

  6. 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)

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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)

  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. System 80+trademark Standard Design: CESSAR design certification

    International Nuclear Information System (INIS)

    1990-01-01

    This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80+trademark Standard Design. This Volume 16 details the application of Human Factors Engineering in the design process

  17. System 80+trademark standard design: CESSAR design certification

    International Nuclear Information System (INIS)

    1990-01-01

    This report has been prepared in support of the industry effort to standardize nuclear plant designs. The documents in this series describe the Combustion Engineering, Inc. System 80+ TM Standard Design

  18. 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.

  19. 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.

  20. 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

  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. System 80+trademark Standard Design: CESSAR design certification

    International Nuclear Information System (INIS)

    1990-01-01

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80 + trademark Standard Design. This volume 10 discusses the Steam and Power Conversion System and Radioactive Waste Management

  3. System 80+trademark Standard Design: CESSAR design certification

    International Nuclear Information System (INIS)

    1990-01-01

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80 + trademark Standard Design. This volume 11 discusses Radiation Protection, Conduct of Operations, and the Initial Test Program

  4. System 80+trademark Standard Design: CESSAR design certification

    International Nuclear Information System (INIS)

    1990-01-01

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80 + trademark Standard Design. This volume 9 discusses Electric Power and Auxiliary Systems

  5. System 80+trademark Standard Design: CESSAR design certification

    International Nuclear Information System (INIS)

    1990-01-01

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describes the Combustion Engineering, Inc. System 80+trademark Standard Design. This Volume 17 provides Appendix A of this report, closure of unresolved and Genetic Safety Issues

  6. System 80+trademark Standard Design: CESSAR design certification

    International Nuclear Information System (INIS)

    1990-01-01

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80 + trademark Standard Design. This volume 8 provides a description of instrumentation and controls

  7. System 80+trademark Standard Design: CESSAR design certification

    International Nuclear Information System (INIS)

    1990-01-01

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80+trademark Standard Design. This Volume 18 provides Appendix B, Probabilistic Risk Assessment

  8. System 80+trademark Standard Design: CESSAR design certification. Volume 16

    International Nuclear Information System (INIS)

    1997-01-01

    This report has been prepared in support of the industry effort to standardize nuclear plant designs. This document describes the Combustion Engineering, Inc. System 80+trademark Standard Design. This volume contain Chapter 18 -- Human Factors Engineering. Topics covered include: design team organization and responsibilities; design goals and design bases; design process and application to human factors engineering; functional task analysis; control room configuration; information presentation and panel layout evaluation; control and monitoring outside the main control room; and verification and validation

  9. 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)

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. State Skill Standards: Housing and Interior Design

    Science.gov (United States)

    Nevada Department of Education, 2008

    2008-01-01

    Meeting the Housing and Interior Design Standards will provide students with skills for personal family life and towards becoming a professional in the interior design field. The mission of Housing and Interior Design education is to prepare students for family life, work life, and careers in the fashion industry by creating opportunities to…

  17. 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)

  18. 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

  19. 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)

  20. DOE Standard: Fire protection design criteria

    International Nuclear Information System (INIS)

    1999-07-01

    The development of this Standard reflects the fact that national consensus standards and other design criteria do not comprehensively or, in some cases, adequately address fire protection issues at DOE facilities. This Standard provides supplemental fire protection guidance applicable to the design and construction of DOE facilities and site features (such as water distribution systems) that are also provided for fire protection. It is intended to be used in conjunction with the applicable building code, National Fire Protection Association (NFPA) Codes and Standards, and any other applicable DOE construction criteria. This Standard replaces certain mandatory fire protection requirements that were formerly in DOE 5480.7A, ''Fire Protection'', and DOE 6430.1A, ''General Design Criteria''. It also contains the fire protection guidelines from two (now canceled) draft standards: ''Glove Box Fire Protection'' and ''Filter Plenum Fire Protection''. (Note: This Standard does not supersede the requirements of DOE 5480.7A and DOE 6430.1A where these DOE Orders are currently applicable under existing contracts.) This Standard, along with the criteria delineated in Section 3, constitutes the basic criteria for satisfying DOE fire and life safety objectives for the design and construction or renovation of DOE facilities

  1. 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

  2. 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

  3. 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)

  4. Assessment of Offshore Wind System Design, Safety, and Operation Standards

    Energy Technology Data Exchange (ETDEWEB)

    Sirnivas, Senu [National Renewable Energy Lab. (NREL), Golden, CO (United States); Musial, Walt [National Renewable Energy Lab. (NREL), Golden, CO (United States); Bailey, Bruce [AWS Trupower LLC, Albany, NY (United States); Filippelli, Matthew [AWS Trupower LLC, Albany, NY (United States)

    2014-01-01

    This report is a deliverable for a project sponsored by the U.S. Department of Energy (DOE) entitled National Offshore Wind Energy Resource and Design Data Campaign -- Analysis and Collaboration (contract number DE-EE0005372; prime contractor -- AWS Truepower). The project objective is to supplement, facilitate, and enhance ongoing multiagency efforts to develop an integrated national offshore wind energy data network. The results of this initiative are intended to 1) produce a comprehensive definition of relevant met-ocean resource assets and needs and design standards, and 2) provide a basis for recommendations for meeting offshore wind energy industry data and design certification requirements.

  5. An operator basis for the Standard Model with an added scalar singlet

    Energy Technology Data Exchange (ETDEWEB)

    Gripaios, Ben [Cavendish Laboratory, J.J. Thomson Avenue, Cambridge (United Kingdom); Sutherland, Dave [Cavendish Laboratory, J.J. Thomson Avenue, Cambridge (United Kingdom); Kavli Institute for Theoretical Physics, UCSB Kohn Hall, Santa Barbara CA (United States)

    2016-08-17

    Motivated by the possible di-gamma resonance at 750 GeV, we present a basis of effective operators for the Standard Model plus a scalar singlet at dimensions 5, 6, and 7. We point out that an earlier list at dimensions 5 and 6 contains two redundant operators at dimension 5.

  6. 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

  7. 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

  8. 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

  9. Design Standards for School Art Facilities

    Science.gov (United States)

    National Art Education Association, 2015

    2015-01-01

    "Design Standards for School Art Facilities" is an invaluable resource for any school or school district looking to build new facilities for the visual arts or renovate existing ones. Discover detailed information about spaces for the breadth of media used in the visual arts. Photographs illustrate all types of features including…

  10. Business School's Performance Management System Standards Design

    Science.gov (United States)

    Azis, Anton Mulyono; Simatupang, Togar M.; Wibisono, Dermawan; Basri, Mursyid Hasan

    2014-01-01

    This paper aims to compare various Performance Management Systems (PMS) for business school in order to find the strengths of each standard as inputs to design new model of PMS. There are many critical aspects and gaps notified for new model to improve performance and even recognized that self evaluation performance management is not well…

  11. 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)

  12. Working group 1B - basis for the standard-liquid pathway

    International Nuclear Information System (INIS)

    Budnitz, R.

    1993-01-01

    This paper presents a summary of the progress made by working group 1B (Basis for the Standard-Liquid Pathway) during the Electric Power Research Institute's EPRI Workshop on the technical basis of EPA HLW Disposal Criteria, March 1993. This group discussed the containment requirements of Environmental Protection Agency (EPA) standard 40 CFR Part 191. This is a major issue when considering the liquid pathway of contamination during the disposal of high-level radioactive wastes. Questions that were raised by this group were (1) What were the problems with Table 1 of 40 CFR Part 191?; (2) What would be fixed with the person-rem alternative that is offered?; and (3) What would be fixed if Table 1 were supplemented with additional columns for other pathways? Other topics that were touched on were the definition of the term open-quotes reasonable expectationsclose quotes and 100,000 year criteria

  13. 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.

  14. 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

  15. 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...

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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

  1. Updated Design Standards and Guidance from the What Works Clearinghouse: Regression Discontinuity Designs and Cluster Designs

    Science.gov (United States)

    Cole, Russell; Deke, John; Seftor, Neil

    2016-01-01

    The What Works Clearinghouse (WWC) maintains design standards to identify rigorous, internally valid education research. As education researchers advance new methodologies, the WWC must revise its standards to include an assessment of the new designs. Recently, the WWC has revised standards for two emerging study designs: regression discontinuity…

  2. A Design Study for Standard Nanofluid Coolants

    International Nuclear Information System (INIS)

    Bang, In Cheol; Heo, Gyun Young

    2007-01-01

    The experimental data for nanofluids in thermal-fluid systems have shown that the new fluids promise to become advanced heat transfer fluids in terms of thermal performance. While enhancing thermal characteristics, the solid-liquid mixtures present an unavoidable disadvantage in terms of pumping cost for economic operation of thermal-fluid systems. In addition, there is a lack of agreement between experimental data provided in the literature. We can find that this issue of nanofluids resembles that of designing new materials. Many nanofluids researchers tend to view the nanofluid field as a highly coupled 'tetrahedro' whose four vertices (performance, properties, structure, and processes) are interconnected to each other. The present design study has a big merit to systemize the nanofluid work and to reduce a lot of trial-error efforts. The present work found that there would be no comprehensible design strategy in developing nanofluids. In this work, the Axiomatic Design (AD) theory is applied to standardize the design of nanofluids in order to bring its practical use forward. According to the Independence Axiom of the AD theory, the excessive couplings between the functional requirements and the parameters of a nanofluid system prevent from meeting the functional goals of the entire system. At a parametric level, the design of a nanofluid system is inherently coupled due to the characteristics of thermal-fluid system; the design parameters physically affect each other sharing sub-level parameters for nanoparticles with making a feedback loop. Even though parts of the nanofluids are naturally coupled, it is possible to reduce and/or eliminate the degree of coupling by help of AD principles. From the perspective of AD, this implies that we are able to ascertain which nanofluid system is better one in the light of functional achievement

  3. 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

  4. 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…

  5. 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

  6. 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

  7. 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

  8. Improved Design Basis for Laterally Loaded Large Diameter Pile

    DEFF Research Database (Denmark)

    Leth, Caspar Thrane

    of up-to-date facilities has been: Equipment for controlling centrifuge tests, data acquisition, preparation of test samples and equipment for and making of lateral load tests. The present research has been narrowed to investigate the static and cyclic behaviour of stiff piles with a diameter of 1-3 m...... in dry sand by use of centrifuge modelling and to compare the findings with the standard p-y curves. It has been chosen to apply piles with an embedment length of 6 to 10 times the diameter of the applied piles. The general static behaviour of monopiles in dry sand has based on centrifuge tests been...... with a capacity of 3.6 MW and placed at water depths beyond 25 m. Different foundation solutions have over the years been applied for offshore turbines, but monopiles are currently the most applied foundation type and are typically with a pile diameter of 4-6 m and applied up to a water depth of 25 m...

  9. 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...

  10. 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...

  11. 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...

  12. Standard mirror fusion reactor design study

    International Nuclear Information System (INIS)

    Moir, R.W.

    1978-01-01

    This report covers the work of the Magnetic Fusion Energy Division's reactor study group during FY 1976 on the standard mirror reactor. The ''standard'' mirror reactor is characterized as a steady state, neutral beam sustained, D-T fusioning plasma confined by a Yin-Yang magnetic mirror field. The physics parameters are obtained from the same physics model that explains the 2XIIB experiment. The model assumes that the drift cyclotron loss cone mode occurs on the boundary of the plasma, and that it is stabilized by warm plasma with negligible energy investment. The result of the study was a workable mirror fusion power plant, steady-state blanket removal made relatively simple by open-ended geometry, and no impurity problem due to the positive plasma potential. The Q (fusion power/injected beam power) turns out to be only 1.1 because of loss out the ends from Coulomb collisions, i.e., classical losses. This low Q resulted in 77% of the gross electrical power being used to power the injectors, thereby causing the net power cost to be high. The low Q stimulated an intensive search for Q-enhancement concepts, resulting in the LLL reactor design effort turning to the field reversal mirror and the tandem mirror, each having Q of order 5

  13. Reduced nicotine product standards for combustible tobacco: building an empirical basis for effective regulation.

    Science.gov (United States)

    Donny, Eric C; Hatsukami, Dorothy K; Benowitz, Neal L; Sved, Alan F; Tidey, Jennifer W; Cassidy, Rachel N

    2014-11-01

    Both the Tobacco Control Act in the U.S. and Article 9 of the Framework Convention on Tobacco Control enable governments to directly address the addictiveness of combustible tobacco by reducing nicotine through product standards. Although nicotine may have some harmful effects, the detrimental health effects of smoked tobacco are primarily due to non-nicotine constituents. Hence, the health effects of nicotine reduction would likely be determined by changes in behavior that result in changes in smoke exposure. Herein, we review the current evidence on nicotine reduction and discuss some of the challenges in establishing the empirical basis for regulatory decisions. To date, research suggests that very low nicotine content cigarettes produce a desirable set of outcomes, including reduced exposure to nicotine, reduced smoking, and reduced dependence, without significant safety concerns. However, much is still unknown, including the effects of gradual versus abrupt changes in nicotine content, effects in vulnerable populations, and impact on youth. A coordinated effort must be made to provide the best possible scientific basis for regulatory decisions. The outcome of this effort may provide the foundation for a novel approach to tobacco control that dramatically reduces the devastating health consequences of smoked tobacco. Copyright © 2014 Elsevier Inc. All rights reserved.

  14. 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

  15. University energy management improvement on basis of standards and digital technologies

    Directory of Open Access Journals (Sweden)

    Novikova Olga

    2018-01-01

    Full Text Available Nowadays to implement the energy management system it is important to fulfill not only the legal requirements but also to follow the set of recommendations prepared by international and national management standards. The purpose of this article is to prepare the concept and methodology for the optimization and improvement of the energy management system (EMS for Universities with implementation of legal requirements and recommendations from international and national management standards with the help of digital technologies. During the research the systematic analysis, complex approach, logical sampling and analogy were used. It is shown that this process should be done with the help of the process-based approach, in accordance with ISO 9001, and energy management ISO 50001. The authors developed the structure of the basic standard of energy management: "Guidelines for the energy management system". It is proved that the involvement of the technical senior students in the project of EMS improvement allows to expand their competencies for new technics and technologies. Cloud service Bitrix24 was chosen for IT-support of the project. During the study, a list of characteristics was used as a basis for creating a query to the technology department of the university. DBMS Microsoft Access was chosen for its creation. In addition, the possible results of initiating a single database containing all the information needed for accounting and control of energy supply were listed. Moreover, the possibility of automated energy management system implementation and its results were considered. The required actions described in this research can be implemented in any University, that will extend energy management to any University worldwide.

  16. System 80+TM standard plant: Design and operations overview

    International Nuclear Information System (INIS)

    Matzie, R.A.; Ritterbusch, S.E.

    1999-01-01

    The System 80+ Standard Plant Design is a 1400 MWe evolutionary Advanced Light Water Reactor (ALWR), designed to meet the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) and the demands of the international market for nuclear power plants which are not only safer but also more economical to maintain and operate. ABB Combustion Engineering Nuclear Power used a defense-in-depth process that (1) adds design margin to basic components to improve performance during normal operation and to decrease the likelihood of an unanticipated transient or an accident, (2) improves the redundancy and diversity of safety systems in order to mitigate design basis accidents and prevent severe accidents, and (3) improves severe accident mitigation capability. This paper describes the most important improved systems and components with emphasis on severe accident prevention and mitigation capability. The improved design features were implemented in an evolutionary manner using proven components. This approach ensures that the plant operates safely and economically, as demonstrated by operating plants in the US and the Republic of Korea. Detailed studies, summarized in this paper, have shown that the System 80+ plant availability is expected to exceed the ALWR requirement of 87% and that the annual operations and maintenance costs are expected to be reduced by $14 million. (author)

  17. Standard High Solids Vessel Design De-inventory Simulant Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burns, Carolyn A.M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Linn, Diana T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smoot, Margaret R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-09-12

    The Hanford Tank 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 be developed that would represent the de-inventory (residual high-density tank solids cleanout) process. Its basis and target characteristics are defined in 24590-WTP-ES-ENG-16-021 and implemented through PNNL Test Plan TP-WTPSP-132 Rev. 1.0. This document describes the de-inventory Newtonian carrier fluid (DNCF) simulant composition that will satisfy the basis requirement to mimic the density (1.18 g/mL ± 0.1 g/mL) and viscosity (2.8 cP ± 0.5 cP) of 5 M NaOH at 25 °C.1 The simulant viscosity changes significantly with temperature. Therefore, various solution compositions may be required, dependent on the test stand process temperature range, to meet these requirements. Table ES.1 provides DNCF compositions at selected temperatures that will meet the density and viscosity specifications as well as the temperature range at which the solution will meet the acceptable viscosity tolerance.

  18. 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 ...

  19. New Buildings Energy Performance Improvement through Incorporation of New Proven Technologies into Standard Designs. Standard Design for TEMF

    National Research Council Canada - National Science Library

    Zhivov, Alexander M

    2004-01-01

    ISSUES: Current Army Standard Designs don't specify potential energy saving and sustainable design opportunities, available energy saving technologies, and technologies resulting in better indoor air quality...

  20. 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

  1. 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

  2. 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)

  3. 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)

  4. 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)

  5. Design and fabrication of NDA standards

    International Nuclear Information System (INIS)

    Long, S.M.; Hsue, S.T.

    1996-01-01

    The Plutonium Facility, TA-55, at Los Alamos National Laboratory is currently producing NDA calibration standards used by various laboratories in the DOE complex. These NIST traceable standards have been produced to calibrate NDA instruments for accountability measurements used for resolving shipper/receiver differences, and for accountability in process residues and process waste. Standards are needed to calibrate various NDA (Non-destructive Assay) instruments such as neutron coincidence counters, gamma-ray counters, and calorimeters. These instruments measure various ranges of nuclear material being produced in the DOE nuclear community. Los Alamos National Laboratory has taken a lead role in fabrication of uranium and plutonium standards, along with other actinides such as neptunium and americium. These standards have been fabricated for several laboratories within the complex. This paper will summarize previous publications detailing the careful planning encompassing components such as precise weighing, destructive analysis, and the use of post fabrication NDA measurements to confirm that the standards meet all preliminary expectations before use in instrument calibration. The paper will also describe the specialized containers, diluents, and the various amount of nuclear materials needed to accommodate the calibration ranges of the instruments

  6. IEEE standard for design qualification of safety systems equipment used in nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    This standard is written to serve as a general standard for qualification of all types of safety systems equipment, mechanical and instrumentation as well as electrical. It also establishes principles and procedures to be followed in preparing specific safety systems equipment standards. Guidance for qualifying specific safety systems equipment may be found in various specific equipment qualification standards that are now available or are being prepared. It is required that safety systems equipment in nuclear power generating stations meet or exceed its performance requirements throughout its installed life. This is accomplished by a disciplined program of design qualification and quality assurance of design, production, installation, maintenance and surveillance. This standard is for the design qualification section of the program only. Design qualification is intended to demonstrate the capability of the equipment design to perform its safety function(s) over the expected range of normal, abnormal, design basis event, post design basis event, and in-service test conditions. Inherent to design qualification is the requirement for demonstration, within limitations afforded by established technical state-of-the-art, that in-service aging throughout the qualified life established for the equipment will not degrade safety systems equipment from its original design condition to the point where it cannot perform its required safety function(s), upon demand. The above requirement reflects the primary role of design qualification to provide reasonable assurance that design- and age-related common failure modes will not occur during performance of safety function(s) under postulated service conditions

  7. 38 CFR 39.22 - Architectural design standards.

    Science.gov (United States)

    2010-07-01

    ... 38 Pensions, Bonuses, and Veterans' Relief 2 2010-07-01 2010-07-01 false Architectural design...-16-10) Standards and Requirements for Project § 39.22 Architectural design standards. The..., Ontario, CA 91761-2816. (a) Architectural and structural requirements—(1) Life Safety Code. Standards must...

  8. 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)

  9. 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

  10. 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

  11. 7 CFR 1717.605 - Design standards, plans and specifications, construction standards, and RUS accepted materials.

    Science.gov (United States)

    2010-01-01

    ..., construction standards, and RUS accepted materials. 1717.605 Section 1717.605 Agriculture Regulations of the... standards, plans and specifications, construction standards, and RUS accepted materials. All borrowers... system design, construction standards, and the use of RUS accepted materials. Borrowers must comply with...

  12. 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.

  13. 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)

  14. 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)

  15. State Skill Standards: Fashion, Textiles and Design

    Science.gov (United States)

    Campbell, Rene Crepaldi; Gaudy, Glenna; Green-Jobe, Victoria; Hatch, Susan; Moen, Julianne; Sheldon, Shannon; Smith, Loree; Chessell, Karen

    2008-01-01

    The mission of Fashion, Textiles and Design Education is to prepare students for family and community life and careers in the fashion industry by creating opportunities to develop the knowledge, skills, attitudes and behaviors needed to: (1) Examine skills needed to effectively manage clothing decisions; (2) Evaluate the use, care and production…

  16. 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)

  17. System 80+{trademark} Standard Design: CESSAR design certification. Volume 3: Amendment I

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-21

    This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 3, in conjunction with Volume 2, provides the design of structures, components, equipment and systems.

  18. 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.

  19. 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)

  20. 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

  1. Beamline standard component designs for the Advanced Photon Source

    International Nuclear Information System (INIS)

    Shu, D.; Barraza, J.; Brite, C.; Chang, J.; Sanchez, T.; Tcheskidov, V.; Kuzay, T.M.

    1994-01-01

    The Advanced Photon Source (APS) has initiated a design standardization and modularization activity for the APS synchrotron radiation beamline components. These standard components are included in components library, sub-components library and experimental station library. This paper briefly describes these standard components using both technical specifications and side view drawings

  2. 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

  3. The System 80+ Standard Plant design control document. Volume 20

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains 2 technical specifications bases as part of Appendix 16 A Tech Spec Bases. They are TS B3.8 Electrical Power Technical Systems Bases and TS B3.9 Refueling Operations Bases. All 3 parts of section 17 (QA) and all 10 parts of section 18 (Human Factors) of the ADM Design and Analysis are contained in this volume. Topics covered in section 17 are: design phase QA; operations phase QA; and design phase reliability assurance. Topics covered by section 18 are: design team organization; design goals; design process; functional task analysis; control room configuration; information presentation; control and monitoring; verification and validation; and review documents

  4. Interior Design Standards in the Secondary FCS Curriculum

    Science.gov (United States)

    Katz, Shana H.; Smith, Bettye P.

    2006-01-01

    This article deals with a study on interior design standards in the secondary FCS curriculum. This study assessed the importance FCS teachers placed on content standards in the interior design curriculum to help determine the amount of time and emphasis to place on the units within the courses. A cover letter and questionnaire were sent…

  5. 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.

  6. 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.

  7. The System 80+ Standard Plant design control document. Volume 23

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains part 16 References and Appendix 19 A Design Alternatives for section 19 (Probabilistic Risk Assessment) of the ADM Design and Analysis. Also covered is section 20 Unresolved Safety Issues of the ADM Design and Analysis. Finally sections 1--6 of the ADM Emergency Operations Guidelines are contained in this volume. Information covered in these sections include: standard post-trip actions; diagnostic actions; reactor trip recovery guideline; LOCA recovery; SG tube rupture recovery

  8. 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)

  9. 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

  10. 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

  11. Basis to demonstrate compliance with the National Emission Standards for Hazardous Air Pollutants for the Stand-off Experiments Range

    Energy Technology Data Exchange (ETDEWEB)

    Michael Sandvig

    2011-01-01

    The purpose of this report is to provide the basis and the documentation to demonstrate general compliance with the National Emission Standard for Hazardous Air Pollutants (NESHAPS) 40 CFR 61 Subpart H, “National Emission Standards for Emissions of Radionuclides Other Than Radon from Department of Energy Facilities,” (the Standard) for outdoor linear accelerator operations at the Idaho National Laboratory (INL) Stand-off Experiments Range (SOX). The intent of this report is to inform and gain acceptance of this methodology from the governmental bodies regulating the INL.

  12. 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.

  13. 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)

  14. 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

  15. Development of Canadian seismic design approach and overview of seismic standards

    Energy Technology Data Exchange (ETDEWEB)

    Usmani, A. [Amec Foster Wheeler, Toronto, ON (Canada); Aziz, T. [TSAziz Consulting Inc., Mississauga, ON (Canada)

    2015-07-01

    Historically the Canadian seismic design approaches have evolved for CANDU® nuclear power plants to ensure that they are designed to withstand a design basis earthquake (DBE) and have margins to meet the safety requirements of beyond DBE (BDBE). While the Canadian approach differs from others, it is comparable and in some cases more conservative. The seismic requirements are captured in five CSA nuclear standards which are kept up to date and incorporate lessons learnt from recent seismic events. This paper describes the evolution of Canadian approach, comparison with others and provides an overview and salient features of CSA seismic standards. (author)

  16. 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.

  17. 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...

  18. Standardization of computer programs - basis of the Czechoslovak library of nuclear codes

    International Nuclear Information System (INIS)

    Gregor, M.

    1987-01-01

    A standardized form of computer code documentation has been established in the CSSR in the field of reactor safety. Structure and content of the documentation are described and codes already subject to this process are mentioned. The formation of a Czechoslovak nuclear code library and facilitated discussion of safety reports containing results of standardized codes are aimed at

  19. System 80+{trademark} Standard Design: CESSAR design certification. Volume 8: Amendment I

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-21

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 8 provides a description of instrumentation and controls.

  20. System 80+{trademark} Standard Design: CESSAR design certification. Volume 11: Amendment I

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-21

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 11 discusses Radiation Protection, Conduct of Operations, and the Initial Test Program.

  1. System 80+{trademark} Standard Design: CESSAR design certification. Volume 9: Amendment I

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-21

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 9 discusses Electric Power and Auxiliary Systems.

  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. The System 80+ Standard Plant design control document. Volume 10

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains Appendices 6A, 6B, and 6C for section 6 (Engineered Safety Features) of the ADM Design and Analysis. Also, parts 1--5 of section 7 (Instrumentation and Control) of the ADM Design and Analysis are covered. The following information is covered in these parts: introduction; reactor protection system; ESF actuation system; system required for safe shutdown; and safety-related display instrumentation

  5. The System 80+ Standard Plant design control document. Volume 17

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains parts 2-7 and appendix 15A for section 15 (Accident Analysis) of the ADM Design and Analysis. Topics covered in these parts are: decrease in heat removal; decrease in RCS flow rate; power distribution anomalies; increase in RCS inventory; decrease in RCS inventory; release of radioactive materials. The appendix covers radiological release models. Also contained here are five technical specifications for section 16 (Technical Specifications) of the ADM Design and Analysis. They are: TS 1.0 Use and Applications; TS 2.0 Safety Limits; TS 3.0 LCO Availability; TS 3.1 Reactivity Control; and TS 3.2 Power Distribution

  6. The System 80+ Standard Plant design control document. Volume 15

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains all five parts of section 12 (Radiation Protection) of the ADM Design and Analysis. Topics covered are: ALARA exposures; radiation sources; radiation protection; dose assessment; and health physics program. All six parts and appendices A and B for section 13 (Conduct of Operations) of the ADM Design and Analysis are also contained in this volume. Topics covered are: organizational structure; training program; emergency planning; review and audit; plant procedures; industrial security; sabotage protection (App 13A); and vital equipment list (App 13B)

  7. The System 80+ Standard Plant design control document. Volume 18

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains the following technical specifications of section 16 (Technical Specifications) of the ADM Design and Analysis: TS 3.3 Instrumentation; TS 3.4 Reactor Coolant System; TS 3.5 Emergency Core Cooling System; TS 3.6 Containment Systems; TS 3.7 Plant Systems; TS 3.8 Electrical Power Systems; TS 3.9 Refueling Operations; TS 4.0 Design Features; TS 5.0 Administrative Controls. Appendix 16 A Tech Spec Bases is also included. It contains the following: TS B2.0 Safety Limits Bases; TS B3.0 LCO Applicability Bases; TS B3.1 Reactivity Control Bases; TS B3.2 Power Distribution Bases

  8. The System 80+ Standard Plant design control document. Volume 1

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers the DCD introduction and contains sections 1 and parts 1--7 of section 2 of the CDM. Parts 1--7 included the following: (2.1) Design of SSC; (2.2) Reactor; (2.3) RCS and connected systems; (2.4) Engineered Safety Features; (2.5) Instrumentation and Control; (2.6) Electric Power; and (2.7) Auxiliary Systems

  9. The System 80+ Standard Plant design control document. Volume 11

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers parts 6 and 7 and appendix 7A for section 7 (Instrumentation and Control) of the ADM Design and Analysis. The topics covered by these are: other systems required for safety; control systems not required by safety; and CMF evaluation of limiting faults. Parts 1--3 of section 8 (Electric Power) of the ADM are also included in this volume. Topics covered by these parts are: introduction; offsite power system; and onsite power system

  10. The System 80+ Standard Plant design control document. Volume 19

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains five technical specification bases that are part of Appendix 16 A of the ADM Design and Analysis. They are: TS B3.3 Instrumentation Bases; TS B3.4 RCS Bases; TS B3.5 ECCS Bases; TS B3.6 Containment Systems Bases; and TS B3.7 Plant Systems Bases

  11. The System 80+ Standard Plant design control document. Volume 21

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains parts 1--10 of section 19 (Probabilistic Risk Assessment) of the ADM Design and Analysis. Topics covered are: methodology; initiating event evaluation; accident sequence determination; data analysis; systems analysis; external events analysis; shutdown risk assessment; accident sequence quantification; and sensitivity analysis. Also included in this volume are Appendix 19.8A Shutdown Risk Assessment and Appendix A to Appendix 19.8A Request for Information

  12. The System 80+ Standard Plant design control document. Volume 2

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers the following information of the CDM: (2.8) Steam and power conversion; (2.9) Radioactive waste management; (2.10) Tech Support Center; (2.11) Initial test program; (2.12) Human factors; and sections 3, 4, and 5. Also covered in this volume are parts 1--6 of section 1 (General Plant Description) of the ADM Design and Analysis

  13. Conceptual basis for prescriptive growth standards from conception to early childhood: present and future.

    Science.gov (United States)

    Uauy, R; Casanello, P; Krause, B; Kuzanovic, J P; Corvalan, C

    2013-09-01

    Healthy growth in utero and after birth is fundamental for lifelong health and wellbeing. The World Health Organization (WHO) recently published standards for healthy growth from birth to 6 years of age; analogous standards for healthy fetal growth are not currently available. Current fetal growth charts in use are not true standards, since they are based on cross-sectional measurements of attained size under conditions that do not accurately reflect normal growth. In most cases, the pregnant populations and environments studied are far from ideal; thus the data are unlikely to reflect optimal fetal growth. A true standard should reflect how fetuses and newborns 'should' grow under ideal environmental conditions. The development of prescriptive intrauterine and newborn growth standards derived from the INTERGROWTH-21(st) Project provides the data that will allow us for the first time to establish what is 'normal' fetal growth. The INTERGROWTH-21(st) study centres provide the data set obtained under pre-established standardised criteria, and details of the methods used are also published. Multicentre study with sites in all major geographical regions of the world using a standard evaluation protocol. These standards will assess risk of abnormal size at birth and serve to evaluate potentially effective interventions to promote optimal growth beyond securing survival. The new normative standards have the potential to impact perinatal and neonatal survival and beyond, particularly in developing countries where fetal growth restriction is most prevalent. They will help us identify intrauterine growth restriction at earlier stages of development, when preventive or corrective strategies might be more effective than at present. These growth standards will take us one step closer to effective action in preventing and potentially reversing abnormal intrauterine growth. Achieving 'optimal' fetal growth requires that we act not only during pregnancy but that we optimize the

  14. The System 80+ Standard Plant design control document. Volume 24

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains sections 7--11 of the ADM Emergency Operations Guidelines. Topics covered are: excess steam demand recovery; loss of all feedwater; loss of offsite power; station blackout recovery; and functional recovery guideline. Appendix A Severe Accident Management Guidelines and Appendix B Lower Mode Operational Guidelines are also included

  15. 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)

  16. Design and industrial production of frequency standards in the USSR

    Science.gov (United States)

    Demidov, Nikolai A.; Uljanov, Adolph A.

    1990-01-01

    Some aspects of research development and production of quantum frequency standards, carried out in QUARTZ Research and Production Association (RPA), Gorky, U.S.S.R., were investigated for the last 25 to 30 years. During this period a number of rubidium and hydrogen frequency standards, based on the active maser, were developed and put into production. The first industrial model of a passive hydrogen maser was designed in the last years. Besides frequency standards for a wide application range, RPA QUARTZ investigates metrological frequency standards--cesium standards with cavity length 1.9 m and hydrogen masers with a flexible storage bulb.

  17. 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

  18. NEG and NIOSH basis for an occupational health standard: 2-diethylaminoethanol

    Energy Technology Data Exchange (ETDEWEB)

    Toren, K.

    1996-05-01

    Health effects associated with occupational exposure to 2-diethylaminoethanol (DEAE) were reviewed as part of the agreement between NIOSH and the Nordic Expert Group for Criteria Documentation of Health Risks from Chemicals to exchange information and expertise in the area of occupational safety and health to provide a scientific basis for the establishment of recommended occupational exposure limits. The occurrence, use, pharmacokinetics, toxicology, immunotoxicity and organ systems, mutagenic, genotoxic, carcinogenic, and reproductive effects of DEAE were reviewed. Three reports of clusters of cases associated with DEAE exposure were described, as were studies examining the dose response relationship of DEAE in humans and experimental animals.

  19. Update of bridge design standards in Alabama for AASHTO LRFD seismic design requirements.

    Science.gov (United States)

    2013-11-01

    The Alabama Department of Transportation (ALDOT) has been required to update their bridge design to the LRFD Bridge Design Specifications. This transition has resulted in changes to the seismic design standards of bridges in the state. Multiple bridg...

  20. Computing tools for implementing standards for single-case designs.

    Science.gov (United States)

    Chen, Li-Ting; Peng, Chao-Ying Joanne; Chen, Ming-E

    2015-11-01

    In the single-case design (SCD) literature, five sets of standards have been formulated and distinguished: design standards, assessment standards, analysis standards, reporting standards, and research synthesis standards. This article reviews computing tools that can assist researchers and practitioners in meeting the analysis standards recommended by the What Works Clearinghouse: Procedures and Standards Handbook-the WWC standards. These tools consist of specialized web-based calculators or downloadable software for SCD data, and algorithms or programs written in Excel, SAS procedures, SPSS commands/Macros, or the R programming language. We aligned these tools with the WWC standards and evaluated them for accuracy and treatment of missing data, using two published data sets. All tools were tested to be accurate. When missing data were present, most tools either gave an error message or conducted analysis based on the available data. Only one program used a single imputation method. This article concludes with suggestions for an inclusive computing tool or environment, additional research on the treatment of missing data, and reasonable and flexible interpretations of the WWC standards. © The Author(s) 2015.

  1. Knowledge about plant is basis for successful cultivation : new international standard handbook on plant physiology

    NARCIS (Netherlands)

    Esch, van H.; Heuvelink, E.; Kierkels, T.

    2015-01-01

    ‘Plant physiology in Greenhouses’ is the new international standard handbook on plant knowledge for the commercial greenhouse grower. It relates the functioning of the plant to the rapid developments in greenhouse cultivation. It is based on a continuing series of plant physiology articles published

  2. Rethinking Space Design Standards Toward Quality Affordable Housing In Malaysia

    Directory of Open Access Journals (Sweden)

    Ishak Nor Haniza

    2016-01-01

    Full Text Available Provision of affordable housing is important to low- and middle-income population. A fit form of house will not only fulfil a basic human need for shelter, but it also contributes to physical and psychological well-being of the occupants. Excellent quality and affordable housing is an indication of a high quality of life. While writings exist on various aspects of the quality of affordable housing in Malaysia, discussion regarding space and design standards has scarcely been given any serious academic attention. Standards concerning residential development usually cover different aspects or stages of the development process. They can include planning standards, design standards, space standards and technical construction standards. The main concern of this paper is on space and design standards specifically. Space standard can be defined as a set of framework which dictates fixed internal space minimums. Meanwhile, design standard indicates design guidelines to ensure the functionality, comfortability and habitability of the house. This paper is concerned exclusively with indoor spaces of a house excluding external circulation spaces and service facilities (in case of strata housing. Its interest is in internal space as an aspect of housing quality. It can be concluded that one of the way forward will be to find the balance between providing adequate minimum spaces for resident satisfactions and having economic values for housing developers. This paper may be used as a valuable reference for authorities and policy makers to better address the best housing space design standards that would benefit not only the occupants, but also the local government and developers alike.

  3. Crew Transportation Technical Standards and Design Evaluation Criteria

    Science.gov (United States)

    Lueders, Kathryn L.; Thomas, Rayelle E. (Compiler)

    2015-01-01

    Crew Transportation Technical Standards and Design Evaluation Criteria contains descriptions of technical, safety, and crew health medical processes and specifications, and the criteria which will be used to evaluate the acceptability of the Commercial Providers' proposed processes and specifications.

  4. 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)

  5. NEURO-SYSTEM OF AIMING AND STABILIZING WITH A REGULATOR ON THE BASIS OF STANDARD MODEL MODEL REFERENCE CONTROLLER

    Directory of Open Access Journals (Sweden)

    B.I. Kuznetsov

    2015-08-01

    Full Text Available The aim of this work is the synthesis of neural network aiming and stabilization system for the special equipment of moving objects with neuro-controller on the basis of standard model and performance comparison of the neural network system with the neural network predictive control. Build a block diagram of the neural network aiming and stabilization system, based on the subject control principle with PD-regulator in the position loop and with neuro-controller on the basis of standard model in the in the velocity loop. The neuro-controller on the basis of standard model Model Reference Controller is synthesized in the MATLAB Neural Network Toolbox and system simulation is performed. The studies show that the transient state variables of the system are oscillatory. Therefore, the neuro-controller with the prediction NN Predictive Controller should be used for aiming and stabilizing system to provide high dynamic characteristics achieved at the cost of higher complexity and computational cost.

  6. 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

  7. 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

  8. International standardization of nuclear reactor designs - the way forward

    International Nuclear Information System (INIS)

    Raetzke, Christian

    2010-01-01

    The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

  9. 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...

  10. Legislatation, standards and awareness of inclusive design in Denmark

    DEFF Research Database (Denmark)

    Colfelt, Solvej; Herriott, Richard

    This presents an overview of the design professions´response to government policies on accessibility and inclusive design. Architectural firms and industrial design firms have different levels of awareness and willingness to implement inclusive design and accessibility objectives. This is driven ...... in part by variation in standards and regulations between industry sectors. Interest groups apparently do not approach accessibility and design-inclusivity objectives in an orchestrated fashion and do not show an awareness of design a means to provide solution.......This presents an overview of the design professions´response to government policies on accessibility and inclusive design. Architectural firms and industrial design firms have different levels of awareness and willingness to implement inclusive design and accessibility objectives. This is driven...

  11. 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

  12. 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.

  13. 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.

  14. Renewable Electricity Standards: Good Practices and Design Considerations

    Energy Technology Data Exchange (ETDEWEB)

    Cox, Sadie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Esterly, Sean [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-01-02

    In widespread use globally, renewable electricity standards (RES) are one of the most widely adopted renewable energy policies and a critical regulatory vehicle to accelerate renewable energy deployment. This policy brief provides an introduction to key RES design elements, lessons from country experience, and support resources to enable more detailed and country-specific RES policy design.

  15. 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

  16. Achieving IT-supported standardized nursing documentation through participatory design

    DEFF Research Database (Denmark)

    Rasmussen, Stine Loft; Lyng, Karen Marie; Jensen, Sanne

    2012-01-01

    that support guideline-based highly structured standard documentation in a large organization with many stakeholders. Applying a participatory design (PD) approach at many organizational levels has involved the stakeholders actively in the design process. Developing a set of design principles has concurrently......In the Capital Region of Denmark a full-scale pilot project on IT-supported nursing documentation is - after running for two months at one full university hospital - showing promising results. In this paper we discuss participatory design as a method to design clinical documentation templates...

  17. A lack of standardization: the basis for the ethical issues surrounding quality and performance reports.

    Science.gov (United States)

    Suchy, Kirsten

    2010-01-01

    Consumers in the United States are taking advantage of the proliferation of publicly available, internet-based performance reports and quality appraisals of health plans, healthcare organizations, hospitals, and physicians to aid in their healthcare decision making. However, these appraisal practices have given rise to controversy and debate over certain distinctive ethical issues. This article advocates a standardized ethical framework to guide current and future development and implementation of performance reports. This framework, which would resolve a number of the major issues, includes the following ethical principles to guide the practice of public reporting on the Internet and facilitate enhanced quality improvement in the healthcare industry: legitimacy, data integrity and quality, transparency, informed understanding, equity, privacy and confidentiality, collaboration, accountability, and evaluation and continuous improvement.

  18. 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

  19. 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...

  20. 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.

  1. Impact of chemistry on Standard High Solids Vessel Design mixing

    Energy Technology Data Exchange (ETDEWEB)

    Poirier, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-02

    The plan for resolving technical issues regarding mixing performance within vessels of the Hanford Waste Treatment Plant Pretreatment Facility directs a chemical impact study to be performed. The vessels involved are those that will process higher (e.g., 5 wt % or more) concentrations of solids. The mixing equipment design for these vessels includes both pulse jet mixers (PJM) and air spargers. This study assesses the impact of feed chemistry on the effectiveness of PJM mixing in the Standard High Solids Vessel Design (SHSVD). The overall purpose of this study is to complement the Properties that Matter document in helping to establish an acceptable physical simulant for full-scale testing. The specific objectives for this study are (1) to identify the relevant properties and behavior of the in-process tank waste that control the performance of the system being tested, (2) to assess the solubility limits of key components that are likely to precipitate or crystallize due to PJM and sparger interaction with the waste feeds, (3) to evaluate the impact of waste chemistry on rheology and agglomeration, (4) to assess the impact of temperature on rheology and agglomeration, (5) to assess the impact of organic compounds on PJM mixing, and (6) to provide the technical basis for using a physical-rheological simulant rather than a physical-rheological-chemical simulant for full-scale vessel testing. Among the conclusions reached are the following: The primary impact of precipitation or crystallization of salts due to interactions between PJMs or spargers and waste feeds is to increase the insoluble solids concentration in the slurries, which will increase the slurry yield stress. Slurry yield stress is a function of pH, ionic strength, insoluble solids concentration, and particle size. Ionic strength and chemical composition can affect particle size. Changes in temperature can affect SHSVD mixing through its effect on properties such as viscosity, yield stress, solubility

  2. Standards for radiation protection instrumentation: design of safety standards and testing procedures

    International Nuclear Information System (INIS)

    Meissner, Frank

    2008-01-01

    This paper describes by means of examples the role of safety standards for radiation protection and the testing and qualification procedures. The development and qualification of radiation protection instrumentation is a significant part of the work of TUV NORD SysTec, an independent expert organisation in Germany. The German Nuclear Safety Standards Commission (KTA) establishes regulations in the field of nuclear safety. The examples presented may be of importance for governments and nuclear safety authorities, for nuclear operators and for manufacturers worldwide. They demonstrate the advantage of standards in the design of radiation protection instrumentation for new power plants, in the upgrade of existing instrumentation to nuclear safety standards or in the application of safety standards to newly developed equipment. Furthermore, they show how authorities may proceed when safety standards for radiation protection instrumentation are not yet established or require actualization. (author)

  3. 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

  4. 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)

  5. 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.)

  6. 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

  7. Differences in safety margins between nuclear and conventional design standards with regards to seismic hazard definition and design criteria

    International Nuclear Information System (INIS)

    Elgohary, M.; Saudy, A.; Orbovic, N.; Dejan, D.

    2006-01-01

    With the surging interest in new build nuclear all over the world and a permanent interest in earthquake resistance of nuclear plants, there is a need to quantify the safety margins in nuclear buildings design in comparison to conventional buildings in order to increase the public confidence in the safety of nuclear power plants. Nuclear (CAN3-N289 series) and conventional (NBCC 2005) seismic standards have different approaches regarding the design of civil structures. The origin of the differences lays in the safety philosophy behind the seismic nuclear and conventional standards. Conventional seismic codes contain the minimal requirement destined primarily to safeguard against major structural failure and loss of life. It doesn't limit damage to a certain acceptable degree or maintain function. Nuclear seismic code requires that structures, systems and components important to safety, withstand the effects of earthquakes. The requirement states that for equipment important to safety, both integrity and functionality should be ascertained. The seismic hazard is generally defined on the basis of the annual probability of exceedence (return period). There is a major difference on the return period and the confidence level for design earthquakes between the conventional and the nuclear seismic standards. The seismic design criteria of conventional structures are based on the use of Force Modification Factors to take into account the energy dissipation by incursion in non-elastic domain and the reserve of strength. The use of such factors to lower intentionally the seismic input is consistent with the safety philosophy of the conventional seismic standard which is the 'non collapse' rather than the integrity and/or the operability of the structures or components. Nuclear seismic standard requires that the structure remain in the elastic domain; energy dissipation by incursion in non-elastic domain is not allowed for design basis earthquake conditions. This is

  8. Design control for standard U.S. EPRTM plants

    International Nuclear Information System (INIS)

    Mathews, Toney A.; Miller, Matthew P.

    2009-01-01

    The U.S. EPR TM design is being reviewed by the U.S. Nuclear Regulatory Commission (NRC) for reference by utility applicants to build and operate EPR TM nuclear reactors in the United States. While the U.S. EPR TM Design Certification and utility Combined License Applications are being reviewed by the NRC, the AREVA-Bechtel Consortium for Engineering Procurement and Construction is proceeding with developing the detailed design. Multiple, parallel regulatory and engineering activities require carefully prepared documents and rigorous design control processes. This paper will review the design control processes used by the AREVA-Bechtel Consortium. Design control must consider the basic design processes required to achieve an integrated, functional design, as well as design change control. Sources of change and the need to keep design bases and licensing bases consistent must be thoroughly understood. An objective of the U.S. EPR TM reactor deployment program for the United States is to achieve maximum standardization of common features of the plant. Such standardization is necessary for economics, speed-of-construction, and operational efficiencies available from a 'fleet' approach to deployment. (author)

  9. Development, improvement and calibration of neutronic reaction rates measurements: elaboration of a standard techniques basis

    International Nuclear Information System (INIS)

    Hudelot, J.P.

    1998-06-01

    In order to improve and to validate the neutronics calculation schemes, perfecting integral measurements of neutronics parameters is necessary. This thesis focuses on the conception, the improvement and the development of neutronics reaction rates measurements, and aims at building a base of standard techniques. Two subjects are discussed. The first one deals with direct measurements by fission chambers. A short presentation of the different usual techniques is given. Then, those last ones are applied through the example of doubling time measurements on the EOLE facility during the MISTRAL 1 experimental programme. Two calibration devices of fission chambers are developed: a thermal column located in the central part of the MINERVE facility, and a calibration cell using a pulsed high flux neutron generator and based on the discrimination of the energy of the neutrons with a time-of-flight method. This second device will soon allow to measure the mass of fission chambers with a precision of about 1 %. Finally, the necessity of those calibrations will be shown through spectral indices measurements in core MISTRAL 1 (UO 2 ) and MISTRAL 2 (MOX) of the EOLE facility. In each case, the associated calculation schemes, performed using the Monte Carlo MCNP code with the ENDF-BV library, will be validated. Concerning the second one, the goal is to develop a method for measuring the modified conversion ratio of 238 U (defined as the ratio of 238 U capture rate to total fission rate) by gamma-ray spectrometry of fuel rods. Within the framework of the MISTRAL 1 and MISTRAL 2 programmes, the measurement device, the experimental results and the spectrometer calibration are described. Furthermore, the MCNP calculations of neutron self-shielding and gamma self-absorption are validated. It is finally shown that measurement uncertainties are better than 1 %. The extension of this technique to future modified conversion ratio measurements for 242 Pu (on MOX rods) and 232 Th (on

  10. Development of standard logic network for PWR NSSS system design

    International Nuclear Information System (INIS)

    Chung, Moon Kyu; Lee, Byeong Ryeong; Lee, Pan Kwon; Lee, Hae Joon; Park, Joon Won; Song, Tae Gil; Kim, Dong Hui; Choi, Hyeon Ho

    1993-05-01

    The self-reliance of NSSS System Design is required not only the design capability to perform the system design but also the management capability to control the resource and time for the Project effectively. The purpose of this study is to develop the simplified standard Logic Network that is scheduled on the time and resource using the PERT/CPM method. That is mainly focused on Ulchin 3, 4 Project. We prepare the management tool of NSSS System Design project. And we can utilize it as a reference tool for the similar project which are complex and long term in a next project. (Author)

  11. Fusion reactor design studies: standard accounts for cost estimates

    International Nuclear Information System (INIS)

    Schulte, S.C.; Willke, T.L.; Young, J.R.

    1978-05-01

    The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed

  12. System 80+{trademark} standard design incorporates radiation protection lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Crom, T.D.; Naugle, C.L. [Duke Engineering & Services, Inc., Charlotte, NC (United States); Turk, R.S. [ABB Combustion Engineering Nuclear Power, Windsor, CT (United States)

    1995-03-01

    Many lessons have been learned from the current generation of nuclear plants in the area of radiation protection. The following paper will outline how the lessons learned have been incorporated into the design and operational philosophy of the System 80+{trademark} Standard Design currently under development by ABB Combustion Engineering (ABB-CE) with support from Duke Engineering and Services, Inc. and Stone and Webster Engineering Corporation in the Balance-of-Plant design. The System 80+{trademark} Standard Design is a complete nuclear power plant for national and international markets, designed in direct response to utility needs for the 1990`s, and scheduled for Nuclear Regulatory Commission (NRC) Design Certification under the new standardization rule (10 CFR Part 52). System 80+{trademark} is a natural extension of System 80{sup R} technology, an evolutionary change based on proven Nuclear Steam Supply System (NSSS) in operation at Palo Verde in Arizona and under construction at Yonggwang in the Republic of Korea. The System 80+{trademark} Containment and much of the Balance of Plant design is based upon Duke Power Company`s Cherokee Plant, which was partially constructed in the late 1970`s, but, was later canceled (due to rapid declined in electrical load growth). The System 80+{trademark} Standard Design meets the requirements given in the Electric Power Research Institute (EPRI) Advanced Light Water Reactor (ALWR) Requirements Document. One of these requirements is to limit the occupational exposure to 100 person-rem/yr. This paper illustrates how this goal can be achieved through the incorporation of lessons learned, innovative design, and the implementation of a common sense approach to operation and maintenances practices.

  13. Design for manufacturability of a VDSM standard cell library

    International Nuclear Information System (INIS)

    Zhou Chong; Zeng Jianping; Chen Lan; Yin Minghui; Zhao Jie

    2012-01-01

    This paper presents a method of designing a 65 nm DFM standard cell library. By reducing the amount of the library largely, the process of optical proximity correction (OPC) becomes more efficient and the need for large storage is reduced. This library is more manufacture-friendly as each cell has been optimized according to the DFM rule and optical simulation. The area penalty is minor compared with traditional library, and the timing, as well as power has a good performance. Furthermore, this library has passed the test from the Technology Design Department of Foundry. The result shows this DFM standard cell library has advantages that improve the yield. (semiconductor integrated circuits)

  14. 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)

  15. 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.

  16. 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

  17. 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)

  18. 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.

  19. 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)

  20. Design standard issues for ITER in-vessel components

    International Nuclear Information System (INIS)

    Majumdar, S.

    1994-01-01

    Unique requirements that must be addressed by a structural design code for the ITER have been summarized. Existing codes such as ASME Section III, or the French RCC-MR were developed primarily for fission reactor out-of-core components and are not directly applicable to the ITER. They may be used either as a guide for developing a design code for the ITER or as interim standards. However, new rules will be needed for handling the irradiation-induced embrittlement problems faced by the ITER blanket components. Design standards developed in the past for the design of fission reactor core components in the United States can be used as guides in this area

  1. Archival-grade optical disc design and international standards

    Science.gov (United States)

    Fujii, Toru; Kojyo, Shinichi; Endo, Akihisa; Kodaira, Takuo; Mori, Fumi; Shimizu, Atsuo

    2015-09-01

    Optical discs currently on the market exhibit large variations in life span among discs, making them unsuitable for certain business applications. To assess and potentially mitigate this problem, we performed accelerated degradation testing under standard ISO conditions, determined the probable disc failure mechanisms, and identified the essential criteria necessary for a stable disc composition. With these criteria as necessary conditions, we analyzed the physical and chemical changes that occur in the disc components, on the basis of which we determined technological measures to reduce these degradation processes. By applying these measures to disc fabrication, we were able to develop highly stable optical discs.

  2. Application of Standardized ITAAC to the APR1400 Design Certification

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Deogji; Kim, Yunho [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Design certification applications (DCAs) submitted pursuant to 10 CFR Part 52 contain two tiers of information, Tier 1 and Tier 2. Tier 1 contains information that is to be certified through rulemaking, and includes the inspections, tests, analyses, and acceptance criteria (ITAAC). Tier 2 contains more detailed information and is the source for information included in Tier1. DC applicants have developed the ITAAC with their own format and contents. This has caused the ineffectiveness of the review process by Nuclear Regulatory Commission (NRC) at the design certification stage. Standardizing the format and content of Tier 1 and ITAAC will achieve greater efficiency and effectiveness for Tier 1 and ITAAC development and NRC review, and for ITAAC implementation and verification. NRC provided the proposed standardized ITAAC for APR1400, and KHNP is considering the application of the proposed ITAAC. The strategy for application of the NRC proposed standardized ITAAC to the APR1400 design certification is now under developing, and a meeting with NRC to review the NRC proposed standardized ITAAC is expected to be held in the near future. In this review process, establishment of industry position on the NRC proposed standardized ITAAC is very important. The support from NEI and US industry is expected to be gained.

  3. Application of Standardized ITAAC to the APR1400 Design Certification

    International Nuclear Information System (INIS)

    Kang, Deogji; Kim, Yunho

    2016-01-01

    Design certification applications (DCAs) submitted pursuant to 10 CFR Part 52 contain two tiers of information, Tier 1 and Tier 2. Tier 1 contains information that is to be certified through rulemaking, and includes the inspections, tests, analyses, and acceptance criteria (ITAAC). Tier 2 contains more detailed information and is the source for information included in Tier1. DC applicants have developed the ITAAC with their own format and contents. This has caused the ineffectiveness of the review process by Nuclear Regulatory Commission (NRC) at the design certification stage. Standardizing the format and content of Tier 1 and ITAAC will achieve greater efficiency and effectiveness for Tier 1 and ITAAC development and NRC review, and for ITAAC implementation and verification. NRC provided the proposed standardized ITAAC for APR1400, and KHNP is considering the application of the proposed ITAAC. The strategy for application of the NRC proposed standardized ITAAC to the APR1400 design certification is now under developing, and a meeting with NRC to review the NRC proposed standardized ITAAC is expected to be held in the near future. In this review process, establishment of industry position on the NRC proposed standardized ITAAC is very important. The support from NEI and US industry is expected to be gained

  4. Codes, standards, and requirements for DOE facilities: natural phenomena design

    International Nuclear Information System (INIS)

    Webb, A.B.

    1985-01-01

    The basic requirements for codes, standards, and requirements are found in DOE Orders 5480.1A, 5480.4, and 6430.1. The type of DOE facility to be built and the hazards which it presents will determine the criteria to be applied for natural phenomena design. Mandatory criteria are established in the DOE orders for certain designs but more often recommended guidance is given. National codes and standards form a great body of experience from which the project engineer may draw. Examples of three kinds of facilities and the applicable codes and standards are discussed. The safety program planning approach to project management used at Westinghouse Hanford is outlined. 5 figures, 2 tables

  5. 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.

  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. SNL/CA Facilities Management Design Standards Manual

    Energy Technology Data Exchange (ETDEWEB)

    Rabb, David [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Clark, Eva [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2014-12-01

    At Sandia National Laboratories in California (SNL/CA), the design, construction, operation, and maintenance of facilities is guided by industry standards, a graded approach, and the systematic analysis of life cycle benefits received for costs incurred. The design of the physical plant must ensure that the facilities are "fit for use," and provide conditions that effectively, efficiently, and safely support current and future mission needs. In addition, SNL/CA applies sustainable design principles, using an integrated whole-building design approach, from site planning to facility design, construction, and operation to ensure building resource efficiency and the health and productivity of occupants. The safety and health of the workforce and the public, any possible effects on the environment, and compliance with building codes take precedence over project issues, such as performance, cost, and schedule.

  8. Reporting Qualitative Research: Standards, Challenges, and Implications for Health Design.

    Science.gov (United States)

    Peditto, Kathryn

    2018-04-01

    This Methods column describes the existing reporting standards for qualitative research, their application to health design research, and the challenges to implementation. Intended for both researchers and practitioners, this article provides multiple perspectives on both reporting and evaluating high-quality qualitative research. Two popular reporting standards exist for reporting qualitative research-the Consolidated Criteria for Reporting Qualitative Research (COREQ) and the Standards for Reporting Qualitative Research (SRQR). Though compiled using similar procedures, they differ in their criteria and the methods to which they apply. Creating and applying reporting criteria is inherently difficult due to the undefined and fluctuating nature of qualitative research when compared to quantitative studies. Qualitative research is expansive and occasionally controversial, spanning many different methods of inquiry and epistemological approaches. A "one-size-fits-all" standard for reporting qualitative research can be restrictive, but COREQ and SRQR both serve as valuable tools for developing responsible qualitative research proposals, effectively communicating research decisions, and evaluating submissions. Ultimately, tailoring a set of standards specific to health design research and its frequently used methods would ensure quality research and aid reviewers in their evaluations.

  9. Views on seismic design standardization of structures, systems and components of nuclear facilities

    International Nuclear Information System (INIS)

    Reddy, G.R.

    2011-01-01

    Structures, Systems and Components (SSCs) of nuclear facilities have to be designed for normal operating loads such as dead weight, pressure, temperature etc., and accidental loads such as earthquakes, floods, extreme, wind air craft impact, explosions etc. Manmade accidents such as aircraft impact, explosions etc., sometimes may be considered as design basis event and sometimes taken care by providing administrative controls. This will not be possible in the case of natural events such as earthquakes, flooding, extreme winds etc. Among natural events earthquakes are considered as most devastating and need to be considered as design basis event which has certain annual frequency specified in design codes. For example nuclear power plants are designed for a seismic event has 10000 year return period. It is generally felt that design of SSCs for earthquake loads is very time consuming and expensive. Conventional seismic design approaches demands for large number of supports for systems and components. This results in large space occupation and in turn creates difficulties for maintenance and in service inspection of systems and components. In addition, complete exercise of design need to be repeated for plants being located at different sites due to different seismic demands. However, advanced seismic response control methods will help to standardize the seismic design meeting the safety and economy. These methods adopt passive, semi active and active devices, and base isolators to control the seismic response. In nuclear industry, it is advisable to go for passive devices to control the seismic responses. Ideally speaking, these methods will make the designs made for normal loads can also satisfy the seismic demand without calling for change in material, geometry, layout etc. in the SSCs. This paper explain the basic ideas of seismic response control methods, demonstrate the effectiveness of control methods through case studies and eventually give the procedure to

  10. 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

  11. 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)

  12. 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)

  13. 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)

  14. 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

  15. 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)

  16. 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

  17. Will the changes proposed to the conceptual framework’s definitions and recognition criteria provide a better basis for the IASB standard setting?

    NARCIS (Netherlands)

    Brouwer, A.; Hoogendoorn, M.; Naarding, E.

    2015-01-01

    In this paper we evaluate the International Accounting Standards Board’s (IASB) efforts, in a discussion paper (DP) of 2013, to develop a new conceptual framework (CF) in the light of its stated ambition to establish a robust and consistent basis for future standard setting, thereby guiding standard

  18. 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.

  19. 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)

  20. 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).

  1. Standardization of green building technologies for environment design

    Directory of Open Access Journals (Sweden)

    Benuzh Andrey

    2016-01-01

    Full Text Available The article describes the structure and field of standardization ISO / TC 205 “Building environment design”, provides examples of green building technologies. The main purpose of the article is to show the interaction between international ISO / TC 205 “Building environment design” and created in Russia in 2016 the Technical Committee of Standardization № 366 “Green technology of the build environment and green innovative products”. Both of these technical committees promote green building technologies for environment design, thereby deal with the negative impact on the environment and the reasons of global warming. Instead of buildings that attempt to suppress and overcome nature, why not design buildings that integrate with the environment, on every possible level? The international standardization work which ISO/TC 205 “Building environment design” performs seeks, in addition to lowering trade barriers for engineering design, to promote and facilitate the design of high performance buildings: higher performing as economic assets for their owners, higher performing as buildings that provide amenable indoor environment for their occupants, and higher performing with respect to resource utilization and environmental impact.

  2. Design Performance Standards for Large Scale Wind Farms

    DEFF Research Database (Denmark)

    Gordon, Mark

    2009-01-01

    of connection into the Eastern Australian power system under the Rules and guidelines set out by AEMC and NEMMCO (AEMO). Where applicable some international practices are also mentioned. Standards are designed to serve as a technical envelope under which wind farm proponents design the plant and maintain...... ongoing technical compliance of the plant during its operational lifetime. This report is designed to provide general technical information for the wind farm connection engineer to be aware of during the process of connection, registration and operation of wind power plants interconnected into the HV TSO......’s network. No special NER Rule has been used in this document, however V30 (year 2009) has been used as the latest reference on some of the topics discussed. Care has been taken to emphasise certain wind farm design and connection issues that could be considered throughout different stages of the wind farm...

  3. Lessons learned from standardized plant design and construction

    International Nuclear Information System (INIS)

    Roche, B.

    1999-01-01

    Following France's hydropower program, Electricite de France began producing electricity with nuclear power during the 60s with units that were all different. In the early 70s, EDF launched an extensive nuclear program which was fueled by the 1973 oil crisis. The particularity of this program is based on the standardization of the design which enables the cost of engineering studies, components and construction to be reduced. As all of the sites presented various conditions, a single design was possible except for the heat sink, connection to the grid and foundations. In order to follow technical progress, the program was divided into several homogeneous series: CP0, CP1 and CP2 for 900 MWe reactors, P4 and P'4 for 1300 MWe reactors and N4 for 1450 MWe reactors. EDF has managed to apply standardization throughout the service life of the plant: all units of the same series are modified in the same manner and with a same batch of modifications. The standardization of operations is also the EDF's rule: technical specifications, safety reports, and safety procedures are normally the same for units belonging to the same series. Nevertheless, when plant design and operations, and heavy maintenance are considered, it becomes increasingly difficult to maintain strict standardization across the board: when examined closely, variations are possible-as regards the chemical specifications of the secondary system, for instance. On the other hand, at the fabrication stage, it is difficult to maintain fabrication procedures and alloy compositions rigorously the same. Standardization offers a tremendous advantage representing 30-40% of construction costs. The main drawback is the risk of generic defects. On the other hand, the risk is rather small owing to the small differences among units. (author)

  4. Leo Spacecraft Charging Design Guidelines: A Proposed NASA Standard

    Science.gov (United States)

    Hillard, G. B.; Ferguson, D. C.

    2004-01-01

    Over the past decade, Low Earth Orbiting (LEO) spacecraft have gradually required ever-increasing power levels. As a rule, this has been accomplished through the use of high voltage systems. Recent failures and anomalies on such spacecraft have been traced to various design practices and materials choices related to the high voltage solar arrays. NASA Glenn has studied these anomalies including plasma chamber testing on arrays similar to those that experienced difficulties on orbit. Many others in the community have been involved in a comprehensive effort to understand the problems and to develop practices to avoid them. The NASA Space Environments and Effects program, recognizing the timeliness of this effort, commissioned and funded a design guidelines document intended to capture the current state of understanding. This document, which was completed in the spring of 2003, has been submitted as a proposed NASA standard. We present here an overview of this document and discuss the effort to develop it as a NASA standard.

  5. American National Standard: nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    1983-01-01

    This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include criteria for acceptance of evaluated nuclear data sets, criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  6. American National Standard nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    A standard is presented which identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  7. 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.

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. A standardized method for beam design in neutron capture therapy

    International Nuclear Information System (INIS)

    Storr, G.J.: Harrington, B.V.

    1993-01-01

    A desirable end point for a given beam design for Neutron Capture Therapy (NCT) should be quantitative description of tumour control probability and normal tissue damage. Achieving this goal will ultimately rely on data from NCT human clinical trials. Traditional descriptions of beam designs have used a variety of assessment methods to quantify proposed or installed beam designs. These methods include measurement and calculation of open-quotes free fieldclose quotes parameters, such as neutron and gamma flux intensities and energy spectra, and figures-of-merit in tissue equivalent phantoms. The authors propose here a standardized method for beam design in NCT. This method would allow all proposed and existing NCT beam facilities to be compared equally. The traditional approach to determining a quantitative description of tumour control probability and normal tissue damage in NCT research may be described by the following path: Beam design → dosimetry → macroscopic effects → microscopic effects. Methods exist that allow neutron and gamma fluxes and energy dependence to be calculated and measured to good accuracy. By using this information and intermediate dosimetric quantities such as kerma factors for neutrons and gammas, macroscopic effect (absorbed dose) in geometries of tissue or tissue-equivalent materials can be calculated. After this stage, for NCT the data begins to become more sparse and in some areas ambiguous. Uncertainties in the Relative Biological Effectiveness (RBE) of some NCT dose components means that beam designs based on assumptions considered valid a few years ago may have to be reassessed. A standard method is therefore useful for comparing different NCT facilities

  18. 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.

  19. 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

  20. 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.

  1. 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.

  2. 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

  3. Regulatory issues resolved through design certification on the System 80+trademark standard plant design

    International Nuclear Information System (INIS)

    Ritterbusch, S.E.; Brinkman, C.B.

    1996-01-01

    The US Nuclear Regulatory Commission (NRC) has completed its review of the System 80+trademark Standard Plant Design, approving advanced design features and closing severe accident licensing issues. Final Design Approval was granted in July 1994. The NRC review was extensive, requiring written responses to over 4,950 questions and formal printing of over 50,000 Safety Analysis Report pages. New safety issues never before addressed in a regulatory atmosphere had to be resolved with detailed analysis and evaluation of design features. the System 80+ review demonstrated that regulatory issues can be firmly resolved only through presentation of a detailed design and completion of a comprehensive regulatory review

  4. 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...

  5. Standardized Elemental Basis for Gas-Turbine Engine Heat Exchangers is the Key Factor for Their Cost Reduction

    Institute of Scientific and Technical Information of China (English)

    Soudarev A.V; Soudarev B.V; Kondratiev V.V; Lazarev M.V

    2001-01-01

    The competitiveness of the small gas turbine units (GTUs) (Ne<300 kW) in the world power market is dependent on both the maintenance expenses and the capital costs of production. Reduction in the maintenance expenditures could be achieved by increasing the plant efficiency. This task could be solved by some methods: increasing the cycle inlet temperature TIT, getting the cycle more complex (use of heat regeneration and compressed air intermediate cooling), cutting the power consumption on heat-stressed parts cooling. Putting the above into effect is linked with introduction of novel structural materials, a sharp increase in the mass-size values and the plant manufacture expenditures, in particular, at provision of its self-regulation.In connection with the above, the development of the combined metal-ceramic airheaters and standardization of the elemental basis of the metal gas-gas heat exchangers will promote reduction in the expenditures of the maintenance and the manufacture of the small-size independent power GTEs.

  6. Side scanner for supermarkets: a new scanner design standard

    Science.gov (United States)

    Cheng, Charles K.; Cheng, J. K.

    1996-09-01

    High speed UPC bar code has become a standard mode of data capture for supermarkets in the US, Europe, and Japan. The influence of the ergonomics community on the design of the scanner is evident. During the past decade the ergonomic issues of cashier in check-outs has led to occupational hand-wrist cumulative trauma disorders, in most cases causing carpal tunnel syndrome, a permanent hand injury. In this paper, the design of a side scanner to resolve the issues is discussed. The complex optical module and the sensor for aforesaid side scanner is described. The ergonomic advantages offer the old counter mounted vertical scanner has been experimentally proved by the industrial funded study at an independent university.

  7. 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

  8. An expanded calibration study of the explicitly correlated CCSD(T)-F12b method using large basis set standard CCSD(T) atomization energies.

    Science.gov (United States)

    Feller, David; Peterson, Kirk A

    2013-08-28

    The effectiveness of the recently developed, explicitly correlated coupled cluster method CCSD(T)-F12b is examined in terms of its ability to reproduce atomization energies derived from complete basis set extrapolations of standard CCSD(T). Most of the standard method findings were obtained with aug-cc-pV7Z or aug-cc-pV8Z basis sets. For a few homonuclear diatomic molecules it was possible to push the basis set to the aug-cc-pV9Z level. F12b calculations were performed with the cc-pVnZ-F12 (n = D, T, Q) basis set sequence and were also extrapolated to the basis set limit using a Schwenke-style, parameterized formula. A systematic bias was observed in the F12b method with the (VTZ-F12/VQZ-F12) basis set combination. This bias resulted in the underestimation of reference values associated with small molecules (valence correlation energies 0.5 E(h)) and an even larger overestimation of atomization energies for bigger systems. Consequently, caution should be exercised in the use of F12b for high accuracy studies. Root mean square and mean absolute deviation error metrics for this basis set combination were comparable to complete basis set values obtained with standard CCSD(T) and the aug-cc-pVDZ through aug-cc-pVQZ basis set sequence. However, the mean signed deviation was an order of magnitude larger. Problems partially due to basis set superposition error were identified with second row compounds which resulted in a weak performance for the smaller VDZ-F12/VTZ-F12 combination of basis sets.

  9. Common basis of establishing safety standards and other safety decision-making levels for different sources of health risk

    International Nuclear Information System (INIS)

    Demin, V.F.

    2002-01-01

    Current approaches in establishing safety standards and other decision-making levels for different sources of health risk are critically analysed. To have a common basis for this decision-making a specific risk index R is recommended. In the common sense R is quantitatively defined as LLE caused by the annual exposure to the risk source considered: R = annual exposure, damage (LLE) from the exposure unit. This common definition is also rewritten in specific forms for a set of different risk sources (ionising radiation, chemical pollutants, etc): for different risk sources the exposure can be measured with different quantities (the probability of death, the exposure dose, etc.). R is relative LLE: LLE in years referred to 1 year under the risk. The dimension of this value is [year/year]. In the statistical sense R is conditionally the share of the year, which is lost due to exposure to a risk source during this year. In this sense R can be called as the relative damage. Really lifetime years are lost after the exposure. R can be in some conditional sense considered as a dimensionless quantity. General safety standards R n for the public and occupational workers have been suggested in terms of this index: R n = 0.0007 and 0.01 accordingly. Secondary safety standards are derived for a number of risk sources (ionising radiation, environmental chemical pollutants, etc). Values of R n are chosen in such a way that to have the secondary radiation BSS being equivalent to the current one's. Other general and derived levels for safety decision-making are also proposed including the de-minimus levels. Their possible dependence on the national or regional health-demographic data (HDD) is considered. Such issues as the ways of the integration and averaging of risk indices considered through the national or regional HDD for different risk sources and the use of non-threshold linear exposure - response relationships for ionising radiation and chemical pollutants are analysed

  10. Evaluation of Standard Concepts Design of Library Interior Physical Environment

    Directory of Open Access Journals (Sweden)

    Debri Harindya Putri

    2018-01-01

    Full Text Available Currently the function of a room is not only used as a shelter, the function of the room itself to be increased as a refreshing or relaxation area for users to follow the development of creativity and technology in the field of design. The comfortable factor becomes the main factor that indicates a successful process of creating a space. No exception library. The nature of library seemed stiff because of its function as a place to read, now can be developed and made into more dynamic with the special design concepts or color patterns used. Libraries can be created a special concept that suits the characteristics of the users themselves. Most users of the library, especially in college libraries are teenagers. Naturally, teenagers like to gather with their friends and we have to facilitate this activity in our library design concept. In addition we can also determine the needs of users through research by questionnaire method. The answers of users can be mapped and drawn conclusions. To explore the research, the author reviewed some literature about library interior design and observed the library of Ma Chung University as a case study. The combined results of the method can be concluded and the discovery of ideal standards of physical environment. So, the library can be made as a comfortable reading environment so as to increased interest in reading behavior and the frequent visits of students in the library

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

    International Nuclear Information System (INIS)

    Schneider, T.C.

    1996-01-01

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

  12. Standard guide for design criteria for plutonium gloveboxes

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This guide defines criteria for the design of glovebox systems to be used for the handling of plutonium in any chemical or physical form or isotopic composition or when mixed with other elements or compounds. Not included in the criteria are systems auxiliary to the glovebox systems such as utilities, ventilation, alarm, and waste disposal. Also not addressed are hot cells or open-face hoods. The scope of this guide excludes specific license requirements relating to provisions for criticality prevention, hazards control, safeguards, packaging, and material handling. Observance of this guide does not relieve the user of the obligation to conform to all federal, state, and local regulations for design and construction of glovebox systems. 1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 1.3 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user...

  13. The renewable portfolio standard: design considerations and an implementation survey

    International Nuclear Information System (INIS)

    Berry, T.; Jaccard, M.

    2001-01-01

    Renewables have social and environmental benefits compared to conventional electricity sources, but are rarely competitive on a strict financial cost basis. This is because conventional sources are sometimes subsidized, their full pollution costs are ignored, and renewables involve newer, higher-cost technologies whose relative costs will fall with commercialization. Governments use several mechanisms to support renewables, including direct financial support (grants, loans), indirect support (R and D, demonstrations), reform of financial costs of conventional sources (subsidy removal, pollution taxes), and the Renewable Portfolio Standard (RPS). The RPS requires a minimum share of electricity from renewable energy sources. Its use is spreading because it maintains an incentive for renewable producers to reduce costs, links the regulated market outcome to an environmental target, and reduces government involvement. Although it is too early to evaluate fully its effectiveness, the survey for this study explored implementation issues in three European countries, nine US states, and Australia, and found the following. The RPS target is usually set to have environmental benefits without causing significant price increases (cost caps are sometimes used). Most jurisdictions limit eligibility to grid-connected, domestic renewables. The RPS is usually applied to producers rather than consumers, and to energy output not capacity. Flexibility mechanisms are desired but a challenge to implement. Administration in the US and Australia is by government with delegation to independent utility regulators, while in Europe it is more the responsibility of government. Everywhere, the RPS is applied alongside other mechanisms of renewables support. (author)

  14. Evaluation of the Swedish biogas standard - basis for a future audit; Utvaerdering av svensk biogasstandard - underlag foer en framtida revision

    Energy Technology Data Exchange (ETDEWEB)

    Svensson, Mattias

    2011-04-15

    This report is intended to constitute a basis for a future revision of the SS 15 54 38, 'Motor fuels - Biogas as fuel for high-speed otto engines.' When it was published in 1999, it was not deemed appropriate to source compressed biomethane from landfill gas, because of its wider range of trace elements, often in relatively high concentrations, such as siloxanes and halogenated hydrocarbons. Siloxanes are also present in the gas from waste water treatment plants, but the assessment was made that upgrading methods available at the time reduced these levels sufficiently. By putting a maximum limit on the nitrogen content, landfill gas was effectively shut out. Technical Development (cryogenic upgrading making it possible to clean landfill gas to biomethane quality; stricter emissions standards for vehicles that has led to a higher level of sophistication in engine and aftertreatment technologies) has now made it necessary to better control the levels of all trace elements. The state of standardization is not satisfactory for CNG in general and renewable CNG (biomethane) in particular. Standardization at the international level is mostly qualitative. Most standards are at the national level and with the exception of Sweden biomethane is only standardized for injection on the natural gas network. A mandate (M/475, 2010) from the European Commission to develop the CEN standards for biomethane fed into the grid, or used directly as renewable CNG, may change this, and the issue will be addressed in a new committee (CEN/TC408 'Project Committee - Biogas for use in transportation and injection into natural gas pipelines'). A variety of trace elements has been discussed for inclusion in the new standards, where countries such as Holland and France stand out as supporters of more stringent and comprehensive regulations. This report's working group considers it like the regulation of substances such as siloxanes and halogenated hydrocarbons will lead

  15. Safety Basis Report

    International Nuclear Information System (INIS)

    R.J. Garrett

    2002-01-01

    As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities

  16. Safety Basis Report

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2002-01-14

    As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities.

  17. 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)

  18. State Energy Efficiency Resource Standards: Design, Status, and Impacts

    Energy Technology Data Exchange (ETDEWEB)

    Steinberg, D.; Zinaman, O.

    2014-05-01

    An energy efficiency resource standard (EERS) is a policy that requires utilities or other entities to achieve a specified amount of energy savings through customer energy efficiency programs within a specified timeframe. EERSs may apply to electricity usage, natural gas usage, or both. This paper provides an overview of the key design features of EERSs for electricity, reviews the variation in design of EERSs across states, and provides an estimate of the amount of savings required by currently specified EERSs in each state. As of December, 2013, 23 states have active and binding EERSs for electricity. We estimate that state EERSs will require annual electricity savings of approximately 8-11% of total projected demand by 2020 in states with EERSs, however the level of savings targeted by the policies varies significantly across states. In addition to the variation in targeted savings, the design of EERSs varies significantly across states leading to differences in the suite of incentives created by the policy, the flexibility of compliance with the policy, the balance of benefits and costs of the policy between producers and consumers, and the certainty with which the policy will drive long-term savings.

  19. 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

  20. 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

  1. 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.

  2. 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

  3. 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…

  4. Design and Characterization of a Photometer-Colorimeter Standard

    Science.gov (United States)

    Eppeldauer, George P.; Rácz, Miklós

    2004-05-01

    A photometer and tristimulus colorimeter has been developed at the National Institute of Standards and Technology (NIST) to realize a color scale. A novel construction was developed to implement the spectral-responsivity-based scale with small uncertainty. The new device can be used as a reference illuminance and luminance meter as well. Temperature-controlled filter combinations, with 5-8 layers in one package, are used to match the responsivity of a silicon tunnel-trap detector to the CIE color-matching functions with small spectral mismatch values (f1'). Design considerations to extend the tunnel-trap detector with replaceable single and double apertures and changeable filter combinations are described. The design and fabrication of the filter packages and the dependence of the f1' values on the thickness of the filter layers are discussed. The colorimeter was characterized for angular, spatial, and spectral responsivity. An improved preamplifier can convert current to voltage in an 11-decade dynamic range with 0.01% uncertainty.

  5. 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...

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. Design of wind turbines for non-standard air density

    DEFF Research Database (Denmark)

    Soraperra, Giusepe

    2005-01-01

    -standard density is intrinsically different, it is impossible to reach the standard rated power at the standard rated speed. Three scenarios are possible (i) to keep the standard rated speed of the turbine by changing the size of the electric generator; (ii) to change the rated speed of the turbine by adopting...... a different pitch angel setting; (iii) adoption of extendeders to the blades can also help in restraining the standard rated power at the standard rated speed for p less than pst. The power curves for the three turbine configurations, each in three different air density conformations, have been calculated...

  12. 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.

  13. 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.

  14. 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

  15. 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...

  16. 75 FR 52860 - Final Airworthiness Design Standards for Acceptance Under the Primary Category Rule; Orlando...

    Science.gov (United States)

    2010-08-30

    ... DEPARTMENT OF TRANSPORTATION Federal Aviation Administration 14 CFR Part 21 Final Airworthiness Design Standards for Acceptance Under the Primary Category Rule; Orlando Helicopter Airways (OHA), Inc... Existence of Proposed Airworthiness Design Standards for Acceptance Under the Primary Category Rule; Orlando...

  17. The STEP standard as an approach for design and prototyping

    OpenAIRE

    Plantec , Alain; Ribaud , Vincent

    1998-01-01

    International audience; STEP is an ISO standard (ISO-10303) for the computer-interpretable representation and exchange of product data. Parts of STEP standardize conceptual structures and usage of information in generic or specific domains. The standardization process of these constructs is an evolutionary approach , which uses generated prototypes at different phases of the process. This paper presents a method for the building of prototype generators, inspired by this standardization proces...

  18. Reference design for the standard mirror hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-05-22

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (/sup 239/Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket.

  19. Reference design for the standard mirror hybrid reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-01-01

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel ( 239 Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket

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

    International Nuclear Information System (INIS)

    Schneider, T.C.

    1995-01-01

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

  1. Mechanical Design of the LHC Standard Half-Cell

    Science.gov (United States)

    Poncet, A.; Brunet, J. C.; Cruikshank, P.; Genet, M.; Parma, V.; Rohmig, P.; Saban, R.; Tavian, L.; Veness, R.; Vlogaert, J.; Williams, L. R.

    1997-05-01

    The LHC Conceptual Design Report issued on 20th October 1995 (CERN/AC/95-05 (LHC) - nicknamed "Yellow Book") introduced significant changes to some fundamental features of the LHC standard half-cell, composed of one quadrupole, 3 dipoles and a set of corrector magnets. A separate cryogenic distribution line was introduced, which was previously inside the main cryostat. The dipole length has been increased from 10 to 15 m and independent powering of the focusing and defocusing quadrupole magnets was chosen. Individual quench protection diodes were introduced in magnets interconnects and many auxiliary bus bars were added to feed in series the various families of correcting superconducting magnets. The various highly intricate basic systems such as: cryostats and cryogenics feeders, superconducting magnets and their electrical feeding and protection, vacuum beam screen and its cooling, support and alignment devices have been redesigned, taking into account the very tight space available. These space constraints are given by the necessity to have maximum integral bending field strength for maximum LHC energy, and the existing LHC tunnel. Finally, cryogenic and vacuum sectorisation have been introduced to reduce downtimes and facilitate commissioning.

  2. 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)

  3. 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).

  4. 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)

  5. 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.

  6. 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

  7. 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.

  8. Materials for Consideration in Standardized Canister Design Activities.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R.; Ilgen, Anastasia Gennadyevna; Enos, David George; Teich-McGoldrick, Stephanie; Hardin, Ernest

    2014-10-01

    This document identifies materials and material mitigation processes that might be used in new designs for standardized canisters for storage, transportation, and disposal of spent nuclear fuel. It also addresses potential corrosion issues with existing dual-purpose canisters (DPCs) that could be addressed in new canister designs. The major potential corrosion risk during storage is stress corrosion cracking of the weld regions on the 304 SS/316 SS canister shell due to deliquescence of chloride salts on the surface. Two approaches are proposed to alleviate this potential risk. First, the existing canister materials (304 and 316 SS) could be used, but the welds mitigated to relieve residual stresses and/or sensitization. Alternatively, more corrosion-resistant steels such as super-austenitic or duplex stainless steels, could be used. Experimental testing is needed to verify that these alternatives would successfully reduce the risk of stress corrosion cracking during fuel storage. For disposal in a geologic repository, the canister will be enclosed in a corrosion-resistant or corrosion-allowance overpack that will provide barrier capability and mechanical strength. The canister shell will no longer have a barrier function and its containment integrity can be ignored. The basket and neutron absorbers within the canister have the important role of limiting the possibility of post-closure criticality. The time period for corrosion is much longer in the post-closure period, and one major unanswered question is whether the basket materials will corrode slowly enough to maintain structural integrity for at least 10,000 years. Whereas there is extensive literature on stainless steels, this evaluation recommends testing of 304 and 316 SS, and more corrosion-resistant steels such as super-austenitic, duplex, and super-duplex stainless steels, at repository-relevant physical and chemical conditions. Both general and localized corrosion testing methods would be used to

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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)

  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.

    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)

  15. 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

  16. 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…

  17. 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

  18. Effect of standards on new equipment design by new international standards and industry restraints

    Science.gov (United States)

    Endelman, Lincoln L.

    1991-01-01

    The use of international standards to further trade is one of the objectives of creating a standard. By having form fit and function compatible the free interchange of manufactured goods can be handled without hindrance. Unfortunately by setting up standards that are peculiar to a particular country or district it is possible to exclude competition from a group of manufacturers. A major effort is now underway to develop international laser standards. In the May I 990 issue of Laser Focus World Donald R. Johnson the director of industrial technology services for the National Institute of Standards and Technology (NIST formerly the National Bureau of Standards) is quoted as follows: " The common means of protectionism has been through certification for the market place. " The article goes on to say " Mr. Johnson expects this tradition to continue and that the new European Community (EC) will demand not just safety standards but performance standards as well. . . . the American laser industry must move very quickly on this issue or risk being left behind the European standards bandwagon. " The article continues laser companies must get involved in the actual standards negotiating process if they are to have a say in future policy. A single set of standards would reduce the need to repeatedly recalibrate products for different national markets. " As a member of ISO TC-72 SC9 I am

  19. 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

  20. IS standards in designing business-to-government collaborations

    NARCIS (Netherlands)

    Flugge, B.

    2010-01-01

    Elaborating the impact of standards on inter-organizational collaborations, inter-organizational studies demonstrated a standard’s positive impact on the collaboration between governmental and business partners. How and under which conditions information systems (IS) standards contribute to the

  1. USL/DBMS NASA/PC R and D project system design standards

    Science.gov (United States)

    Dominick, Wayne D. (Editor); Moreau, Dennis R.

    1984-01-01

    A set of system design standards intended to assure the completeness and quality of designs developed for PC research and development projects is established. The standards presented address the areas of problem definition, initial design plan, design specification, and re-evaluation.

  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. 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

  5. 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.

  6. Eulerian Circles (Venn Diagrams) as model for modern economy education on the basis of Russian professional standards

    Science.gov (United States)

    Sharonov, M. A.; Sharonova, O. V.; Sharonova, V. P.

    2018-03-01

    The article is an attempt to create a model built using Eulerian circles (Venn diagrams) to illustrate the methodological impact of recent Federal Law 283-FZ “On the independent evaluation of qualifications” and new Federal State Educational Standards of higher education of generation 3++ on educational process in Russia. In modern economic conditions, the ability to correctly assess the role of professional standards, as a matter of fact, some set, the degree of intersection with the approximate basic educational program and the Federal State Educational Standards becomes an important factor on which in the future will depend not only the demand of graduates in the labor market, but also the possibility of passing the professional and public accreditation of the proposed program.

  7. Standards in capstone design courses and the engineering curriculum.

    Science.gov (United States)

    Goldberg, Jay

    2012-01-01

    In 1990, I was the group leader of Research and Development for Surgitek, a urological device manufacturer in Racine, Wisconsin, which was the market leader in ureteral stents. I recognized the need for an industry standard for ureteral stents, felt that taking a leadership role in the development of a new standard would be a good professional development activity, and volunteered to serve as the chair of the American Society for Testing and Materials (ASTM, now known as ASTM International) Task Group on Ureteral Stents. After several years, the standard was approved, and I continue to serve as the chair of the ASTM Subcommittee on Urological Materials and Devices. This experience helped me understand the value of industry standards, expanded my professional network and visibility within the medical device industry, and helped develop my leadership skills.

  8. 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.

  9. CytometryML: a data standard which has been designed to interface with other standards

    Science.gov (United States)

    Leif, Robert C.

    2007-02-01

    Because of the differences in the requirements, needs, and past histories including existing standards of the creating organizations, a single encompassing cytology-pathology standard will not, in the near future, replace the multiple existing or under development standards. Except for DICOM and FCS, these standardization efforts are all based on XML. CytometryML is a collection of XML schemas, which are based on the Digital Imaging and Communications in Medicine (DICOM) and Flow Cytometry Standard (FCS) datatypes. The CytometryML schemas contain attributes that link them to the DICOM standard and FCS. Interoperability with DICOM has been facilitated by, wherever reasonable, limiting the difference between CytometryML and the previous standards to syntax. In order to permit the Resource Description Framework, RDF, to reference the CytometryML datatypes, id attributes have been added to many CytometryML elements. The Laboratory Digital Imaging Project (LDIP) Data Exchange Specification and the Flowcyt standards development effort employ RDF syntax. Documentation from DICOM has been reused in CytometryML. The unity of analytical cytology was demonstrated by deriving a microscope type and a flow cytometer type from a generic cytometry instrument type. The feasibility of incorporating the Flowcyt gating schemas into CytometryML has been demonstrated. CytometryML is being extended to include many of the new DICOM Working Group 26 datatypes, which describe patients, specimens, and analytes. In situations where multiple standards are being created, interoperability can be facilitated by employing datatypes based on a common set of semantics and building in links to standards that employ different syntax.

  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. 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

  12. Identification of Producibility Inhibitors in Standard NAVSEA Ship Design Practices

    National Research Council Canada - National Science Library

    Wilkins, James R; Kraine, Gilbert L; LaChance, George M; Burnett, Victor R

    1993-01-01

    .... This report provides the results of interviews with and questionnaires returned from various shipyards, ship design agents and supervisors of shipbuilding who were asked to identify producibility...

  13. 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

  14. 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

  15. 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.

  16. 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

  17. Communication Standards Suitable for MES Systems Designed for SMES

    Directory of Open Access Journals (Sweden)

    Ondřej WINKLER

    2009-06-01

    Full Text Available A MES (Manufacturing Execution System is a system that large companies use to measure or control critical production activities. Individual parts of this system can be used as a model for development of tools suitable for small and medium enterprises (SMEs. It can demonstrate how manufacturing systems works and which communication standards are usable. The communication standards enable data acquisition, data processing, storing and their conversion for any production report. Furthermore, interfaces for web clients allow displaying the raw or transformed data as a text or a graphic chart. The development of tools for an access to the data retrieved from MES software is described in this paper.

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. Designing RF control subsystems using the VXIbus standard

    International Nuclear Information System (INIS)

    Stepp, J.D.; Vong, F.C.; Bridges, J.F.

    1993-01-01

    Various components are being designed to control the RF system of the 7-GeV Advanced Photon Source (APS). The associated control electronics (phase shifters, amplitude modulators, phase detectors, automatic tuning control, and local feedback control) are designed as modular cards with multiple channels for ease of replacement as well as for compact design. Various specifications of the VXIbus are listed and the method used to simplify the design of the control subsystem is shown. A commercial VXI interface board was used to speed the design cycle. Required manpower and actual task times are included. A discussion of the computer architecture and software development of the device drivers which allowed computer control from a VME processor located in a remote crate operating under the Experimental Physics and Industrial Controls Software (EPICS) program is also presented

  6. Windmills by Design: Purposeful Curriculum Design to Meet Next Generation Science Standards in a 9-12 Physics Classroom

    Science.gov (United States)

    Concannon, James; Brown, Patrick L.

    2017-01-01

    The "Next Generation Science Standards" (NGSS) challenges science teachers to think beyond specific content standards when considering how to design and implement curriculum. This lesson, "Windmills by Design," is an insightful lesson in how science teachers can create and implement a cross-cutting lesson to teach the concepts…

  7. 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.

  8. Designing Standardized Patient Assessments to Measure SBIRT Skills

    Science.gov (United States)

    Wamsley, Maria A.; Julian, Katherine A.; O'Sullivan, Patricia; Satterfield, Jason M.; Satre, Derek D.; McCance-Katz, Elinore; Batki, Steven L.

    2013-01-01

    Objectives: Resident physicians report insufficient experience caring for patients with substance use disorders (SUDs). Resident training in Screening, Brief Intervention, and Referral to Treatment (SBIRT) has been recommended. We describe the development of a standardized patient (SP) assessment to measure SBIRT skills, resident perceptions of…

  9. 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

  10. Military Housing: Status of the Services' Implementation of the Current Barracks Design Standard

    National Research Council Canada - National Science Library

    1999-01-01

    .... In November 1995, DOD adopted a new barracks construction standard, referred to as the 1+1 design standard, that called for more space and increased privacy in new barracks for service members permanently assigned to an installation...

  11. Standards and interdisciplinary treatment of boxing injuries of the head in professional boxing on the basis of an IBF World Championship Fight.

    Science.gov (United States)

    Dragu, Adrian; Unglaub, Frank; Radomirovic, Sinisa; Schnürer, Stefan; Wagner, Walter; Horch, Raymund E; Hell, Berthold

    2010-12-01

    Boxing injuries are well known in hobby boxing as well as in professional boxing. Especially in professional boxing it is of great importance to implement and follow prevention-, diagnosis- and therapy-standards in order to prevent or at least to minimize injuries of the athlete. The utmost aim would be to establish international prevention-, diagnosis- and therapy-standards for boxing injuries in professional boxing. However, this aim is on a short run unrealistic, as there are too many different professional boxing organisations with different regulations. A realistic short term aim would be to develop a national standard in order to unify the management and medical treatment of boxing injuries in professional boxing. We present the management and interdisciplinary treatment of a professional boxer with a bilateral open fracture of the mandible during a middle weight IBF World Championship Fight. On the basis of this case we want to present and discuss the possibilities of an interdisciplinary and successful medical treatment. In order to prevent or minimize boxing injuries of professional boxers, annual MRI-Scans of the head and neck have to be performed as prevention standard. Furthermore, neurocognitive tests must be performed on a regular basis. Boxing injuries in professional boxing need an interdisciplinary, unbiased and complex analysis directly at the boxing ring. The treatment of the injuries should be only performed in medical centres and thus under constant parameters. The needed qualifications must be learned in mandatory national licence courses of boxing physicians, referees and promoters.

  12. 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)

  13. 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

  14. History of Road Design Standards in LADOTD : Research Project Capsule

    Science.gov (United States)

    2012-10-01

    According to the Louisiana Statewide Transportation Plan, Louisianas highway : network is comprised of over 60,000 miles, of which over 16,000 miles are : maintained by the state. The roadways were designed and constructed : according to the desig...

  15. SOA and Web services interface design principles, techniques, and standards

    CERN Document Server

    Bean, James

    2009-01-01

    SOA offers solutions to the most intractable business problems faced by every enterprise, but getting the SOA service interface right requires the practical design knowledge this book uniquely delivers

  16. 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

  17. 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.

  18. 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)

  19. An investigation into standards in sustainable design and manufacture

    OpenAIRE

    Plant, AVC; Harrison, DJ; Griffiths, BJ; Lam, B

    2009-01-01

    This paper reports upon the application of standards to reduce the negative environmental impacts of manufacturing through product lifecycle planning and closed loop production. By eliminating waste and retaining the energy embodied within materials and components, manufacturing can become more sustainable from both ecological and financial perspectives. Energy consumption and the associated carbon pollution can thus be minimised. Environmental Management System implementation is also cons...

  20. Design and Characterization of Vertical Mesh Capacitors in Standard CMOS

    DEFF Research Database (Denmark)

    Christensen, Kåre Tais

    2001-01-01

    This paper shows how good RF capacitors can be made in a standard digital CMOS process. The capacitors which are also well suited for binary weighted switched capacitor banks show very good RF performance: Q-values of 57 at 4.0 GHz, a density of 0.27 fF/μ2, 2.2 μm wide shielded unit capacitors, 6...

  1. Using a standards committee to design practical procedure system improvements

    International Nuclear Information System (INIS)

    Grider, D.A.; Plung, D.

    1993-01-01

    In the post-Three Mile Island (TMI) environment, numerous reports have been issued on how to improve the quality of procedures used at government and commercial nuclear facilities. The studies tend to be long on what is wrong with existing procedures and short on practical directions on how to fix those faults. Few of these studies have been conducted by practitioners with full-time procedure-managing or procedure writing experience. None of these studies go into detail on how to improve the procedure system itself. Over the last 10 yr, various nuclear facilities within the US Department of Energy (DOE) have carried out individual programs to develop procedures that meet post-TMI standards. However, ∼2 yr ago, DOE formed a Procedures Standards Committee to advise DOE in developing a set of post-TMI guidelines that could be consistently applied throughout all DOE nuclear facilities. The committee has achieved not only its original mission by producing a series of integrated guidance documents but has also evolved a systems approach to procedures management that sets new standards for procedure quality and efficiency. As members of this committee, the authors want to describe what has made the group's approach so successful. The lessons learned may be translatable to a wide range of government and commercial industry procedure programs

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. Critical analysis of accident scenario and consequences modelling applied to light-water reactor power plants for accident categories beyond the design basis accident (DBA)

    International Nuclear Information System (INIS)

    Brofferio, C.; Cagnetti, P.; Ferrara, V.; Manilia, E.; Pietrangeli, G.; Sennis, C.

    1985-01-01

    A critical analysis and sensitivity study of the modelling of accident scenarios and environmental consequences are presented, for light-water reactor accident categories beyond the standard design-basis-accident category. The first chapter, on ''source term'' deals with the release of fission products from a damaged core inventory and their migration within the primary circuit and the reactor containment. Particular attention is given to the influence of engineering safeguards intervention and of the chemical forms of the released fission products. The second chapter deals with their release to the atmosphere, transport and wet or dry deposition, outlining relevant partial effects and confronting short-duration or prolonged releases. The third chapter presents a variability analysis, for environmental contamination levels, for two extreme hypothetical scenarios, evidencing the importance of plume rise. A numerical plume rise model is outlined

  8. 24 CFR 941.203 - Design and construction standards.

    Science.gov (United States)

    2010-04-01

    ... national building code, such as Uniform Building Code, Council of American Building Officials Code, or Building Officials Conference of America Code; (2) Applicable State and local laws, codes, ordinances, and... intended to serve. Building design and construction shall strive to encourage in residents a proprietary...

  9. Designation Of The Standard Stratotype And Age Of Nsukka ...

    African Journals Online (AJOL)

    The type locality of Nsukka Formation was designated north of Nsukka in an area along Nadu River, without any formal description of the type section. Field mapping of the type locality was undertaken with the intention to formally define the type section. The thickest and best exposed is a 40 meter thick, stream section at ...

  10. Recent developments for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, N.; Nakamuria, K.; Morishita, M.; Shibamoto, H.; Nagashima, H.; Inoue, K.

    2005-01-01

    For realization of reliable and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute(JNC) and Japan Atomic Power Company(JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts through evaluation of their structural integrity, the research and development of 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)' is recognized as an essential theme. FDS focuses on particular failure modes of FRs such as ratchet deformation and creep fatigue damages due to cyclic thermal loads. To evaluate these modes, three main developments are in progress. One is 'Refinement of Failure Criteria' for particular modes of FRs. Next is development of 'Guidelines for Inelastic Design Analysis' in order to predict elastic plastic and creep behaviors. Furthermore, efforts are being made toward preparing 'Guidelines for Thermal Load Modeling' for FR component design where thermal loads are dominant. These studies were performed under the sponsorship of the Ministry of Economy, Trade and Industry of Japanese government. (authors)

  11. 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.

  12. 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

  13. The development of an ergonomics standard for the design of operator interfaces

    International Nuclear Information System (INIS)

    Kirwan, B.; Reed, J.; Litherland, M.

    1990-01-01

    BNFL has realised the need to take a consistent approach to the ergonomic design of operator interfaces. Towards this aim, a design standard document has been produced under the direction of a principal design engineer based on the key ergonomics aspects of plant design. The standard was requested by the designers, and the original standard was produced by ergonomists working on BNFL projects. This standard was then reviewed by a large number of key design and operations personnel, and a series of multidisciplinary meetings produced the final version. The standard contains six sections (ergonomics requirements for the design of Control Rooms, Consoles and Panels Design, Labelling, VDU Systems, Alarm Systems and Colour Coding) containing approximately 180 guidelines in text format or supplemented by diagrams and tables. Each guideline is classified as either mandatory or advisory. A high proportion of effort concentrated on making the document usable by designers. The standard is not intended to be fully comprehensive, since the range of possible variations in the designs of interfaces makes such a task intractable at this stage. However, the document does ensure that account is taken of ergonomics throughout the design phase, and particularly in the early phases whilst design change is still cost-effective, and that designers are aware of the important issues and principles. (author)

  14. Russian standards and design practice of ensuring NPP reliability under severe external loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Birbraer, A N [St. Petersburg Research and Design Institute Atomenergoproject, St. Petersburt (Russian Federation)

    1993-07-01

    Russian Standards and design practice of ensuring NPP reliability under severe external loading conditions are described. The main attention is paid to the seismic design requirements. Explosions, aircraft impact, and tornado are briefly examined too (author)

  15. Russian standards and design practice of ensuring NPP reliability under severe external loading conditions

    International Nuclear Information System (INIS)

    Birbraer, A.N.

    1993-01-01

    Russian Standards and design practice of ensuring NPP reliability under severe external loading conditions are described. The main attention is paid to the seismic design requirements. Explosions, aircraft impact, and tornado are briefly examined too (author)

  16. Review of international standards related to the design for control rooms on nuclear power plants

    International Nuclear Information System (INIS)

    Kitamura, Masashi; Yoshikawa, Hidekazu; Fujita, Yushi

    2005-01-01

    The improvement of Human-Machine Interface (HMI) design for control rooms on nuclear power plants (NPP) has been accomplished world wide, especially after the TMI-2 accident. The design process and guidelines are standardized in IEC60964 and supplemental standards as international standard. However, technological update is required due to the increased use of computerized control and monitoring equipment and systems in control rooms on NPP in recent years. Standards are becoming more important for computerized control rooms because there is more freedom to design than conventional hardware based system. For computerized control rooms, standards for hardware and software of HMI systems should be also considered. Standards and guidelines for computerized control rooms on NPP have been developed recently in each body such as IEC, ISO, and IEEE etc. Therefore, reviewing these standards and guidelines related to control rooms design of NPP can be useful not only for revision of the international standards such as IEC60964, but also for users of the standards and guidelines. In this paper, we reviewed the international standards related to the design for control rooms, in the two aspects of HMI design and hardware and software design, considering the undergoing revision work and their application. (author)

  17. A preliminary study on the design in architecture of nuclear and radiation safety standard system

    International Nuclear Information System (INIS)

    Song Dahu; Zhang Chi; Yang Lili; Li Bin; Liu Yingwei; An Hongzhen; Gao Siyi; Liu Ting; Meng De

    2014-01-01

    The connotation and function of nuclear and radiation safety standards are analyzed, and their relationships with the relevant laws and regulations are discussed in the paper. Some suggestions and blue print of overall architecture to build nuclear and radiation safety standard system are proposed, on the basis of researching the application status quo, existing problems and needs for nuclear and radiation safety standards in China. This work is a beneficial exploration and attempt to establish China's nuclear and radiation safety standards. (authors)

  18. 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).

  19. Design of nuclear spectroscopy electronics based on the EUROsystem standard

    International Nuclear Information System (INIS)

    Pahor, J.

    1987-11-01

    The development of modular nuclear instruments in the EUROcard system has been continued by the design of a staircase generator; such an instrument is needed for accurate testing of spectroscopy amplifiers. The generator provides very precisely defined steps with the period between them varying from 8 to 50 microseconds, the output voltage continuously variable from 2 to 5 V, and attenuated ion steps between 1 and 100. Great care was taken that individual steps exhibit no overshot, thus providing a very clean signal to the input of a tested amplifier. The circuitry is kept simple, and it can be easily constructed in any electronics laboratory. Figs

  20. 36 CFR 223.38 - Standards for road design and construction.

    Science.gov (United States)

    2010-07-01

    ... Provisions § 223.38 Standards for road design and construction. Road construction authorized under timber... 36 Parks, Forests, and Public Property 2 2010-07-01 2010-07-01 false Standards for road design and construction. 223.38 Section 223.38 Parks, Forests, and Public Property FOREST SERVICE, DEPARTMENT OF...

  1. 36 CFR 1281.4 - What are the architectural and design standards for Presidential libraries?

    Science.gov (United States)

    2010-07-01

    ... and design standards for Presidential libraries? 1281.4 Section 1281.4 Parks, Forests, and Public Property NATIONAL ARCHIVES AND RECORDS ADMINISTRATION NARA FACILITIES PRESIDENTIAL LIBRARY FACILITIES § 1281.4 What are the architectural and design standards for Presidential libraries? The Archivist is...

  2. Design and analysis of control charts for standard deviation with estimated parameters

    NARCIS (Netherlands)

    Schoonhoven, M.; Riaz, M.; Does, R.J.M.M.

    2011-01-01

    This paper concerns the design and analysis of the standard deviation control chart with estimated limits. We consider an extensive range of statistics to estimate the in-control standard deviation (Phase I) and design the control chart for real-time process monitoring (Phase II) by determining the

  3. 76 FR 72097 - Air Quality Designations for the 2008 Lead (Pb) National Ambient Air Quality Standards

    Science.gov (United States)

    2011-11-22

    ... Air Quality Designations for the 2008 Lead (Pb) National Ambient Air Quality Standards AGENCY: Environmental Protection Agency (EPA). ACTION: Final rule. SUMMARY: This rule establishes air quality designations for most areas in the United States for the 2008 lead (Pb) National Ambient Air Quality Standards...

  4. 77 FR 30087 - Air Quality Designations for the 2008 Ozone National Ambient Air Quality Standards

    Science.gov (United States)

    2012-05-21

    ... and 81 Air Quality Designations for the 2008 Ozone National Ambient Air Quality Standards; Implementation of the 2008 National Ambient Air Quality Standards for Ozone: Nonattainment Area Classifications...-9668-2] RIN 2060-AP37 Air Quality Designations for the 2008 Ozone National Ambient Air Quality...

  5. Standard and alternative landfill capping design in Germany

    International Nuclear Information System (INIS)

    Simon, Franz-Georg; Mueller, Werner W.

    2004-01-01

    Engineered capping systems are in most cases an indispensable and often the only efficient component required by the long-term safety concept for landfills, mine tailings tips and contaminated land. In Germany the composite liner is the main component of standard landfill cappings for municipal and hazardous waste landfills and the compacted clay liner (CCL) for landfills for inert or low-contamination waste. The composite liner is a technically highly effective but very expensive system. Research and experience has given rise to concern about the proper long-term performance of a conventional single CCL as a landfill capping. Therefore, alternative capping systems are discussed and applied for landfills and for the containment of contaminated sites. This paper gives an overview on various alternative engineered cappings and suitable systems for capping reflecting the state of the art and the expert view in Germany. According to the European Council Directive on the landfill of waste an impermeable mineral layer is recommended for the surface sealing of non-hazardous landfills and a composition of artificial sealing liner and impermeable mineral layer for hazardous landfills. In both cases a drainage layer thickness of at least 0.5 m is suggested. These recommendations should be interpreted flexibly and to some extent modified in the light of the experience and results presented in this paper

  6. C. F. Braun. Standard turbine island design, safety analysis report

    International Nuclear Information System (INIS)

    1974-01-01

    A standard turbine island used with a BWR is described. It consists of the turbine-generator; steam system; condensate storage, cleanup, and transfer systems; control and instrumentation; water treatment plant; make-up demineralizer; potable and waste water systems; and a compressed air system. The turbine-generator is a tandem-compound nuclear-type turbine with one double-flow high-pressure section and a six-flow low-pressure section in three double-flow low-pressure casings. The turbine is direct connected to an 1800 rpm synchronous a-c generator. A combined moisture separator and two-stage reheater is provided. The main steam system delivers the steam generated in a BWR to the main turbine stop valves. The condensate system maintains proper water inventory. Protective features prevent loss of the system due to electrical failure of a component and isolates faults to ensure continuity of a power supply from alternate sources. (U.S.)

  7. 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)

  8. Exemplary Design Envelope Specification for Standard Modular Hydropower Technology

    Energy Technology Data Exchange (ETDEWEB)

    Witt, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Brennan T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Tsakiris, Achilleas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Papanicolaou, Thanos [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Pries, Jason L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burress, Timothy A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Stewart, Kevin M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lee, Kyutae [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DeNeale, Scott T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevelhimer, Mark S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pracheil, Brenda M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McManamay, Ryan A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Kutz, Benjamin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Curd, Shelaine L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bishop, Norm [Knight Piesold, Denver, CO (United States); Welch, Tim [U.S. Department of Energy (DOE), Washington, DC (United States). Office of Energy Efficiency and Renewable Energy (EERE); Rabon, Daniel [U.S. Department of Energy (DOE), Washington, DC (United States). Office of Energy Efficiency and Renewable Energy (EERE); Fernandez, Alisha [McKeown and Associates, Moberly, MO (United States)

    2017-02-01

    Hydropower is an established, affordable renewable energy generation technology supplying nearly 18% of the electricity consumed globally. A hydropower facility interacts continuously with the surrounding water resource environment, causing alterations of varying magnitude in the natural flow of water, energy, fish, sediment, and recreation upstream and downstream. A universal challenge in facility design is balancing the extraction of useful energy and power system services from a stream with the need to maintain ecosystem processes and natural environmental function. On one hand, hydroelectric power is a carbon-free, renewable, and flexible asset to the power system. On the other, the disruption of longitudinal connectivity and the artificial barrier to aquatic movement created by hydraulic structures can produce negative impacts that stress fresh water environments. The growing need for carbon-free, reliable, efficient distributed energy sources suggests there is significant potential for hydropower projects that can deploy with low installed costs, enhanced ecosystem service offerings, and minimal disruptions of the stream environment.

  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. Monitoring of dams according to German standard DIN 19 700 on the basis of a GIS based expert system; Stauanlagenueberwachung nach DIN 19 700 in Verbindung mit Expertensystemen auf GIS-Basis

    Energy Technology Data Exchange (ETDEWEB)

    Kast, Karl; Saucke, Ulrich; Elsaesser, Ralph [Dr.-Ing. Karl Kast + Partner (GbR) Ingenieurgemeinschaft fuer Umwelt- und Geotechnik, Ettlingen (Germany)

    2010-07-01

    The German standard DIN 19 700 stipulates regular safety reports assessing dams and their operation as part of the continuous dam monitoring scheme. Experience has shown that a GIS based dam monitoring system exhibits evident advantages regarding a continuous assessment. Besides and easy access to all relevant information grouped by themes, it offers a graphic analysis of measuring data and their display in the GIS system. In addition, concerning relevant failure mechanism, individual modules of a GIS based expert system are designed for the purpose of assessing the safety of the dam in the event of uncommon incidents. (orig.)

  20. 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

  1. 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)

  2. 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.

  3. 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.

  4. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

  5. Combining economic and social goals in the design of production systems by using ergonomics standards

    NARCIS (Netherlands)

    J. Dul (Jan); H.J. de Vries (Henk); S. Verschoof (Sandra); W. Eveleens (Wietske); A. Feilzer (Albert)

    2004-01-01

    textabstractIn designing of production systems, economic and social goals can be combined, if ergonomics is integrated into the design process. More than 50 years of ergonomics research and practice have resulted in a large number of ergonomics standards for designing physical and organizational

  6. 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.

  7. 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

  8. 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

  9. Ergonomics standards and guidelines for computer workstation design and the impact on users' health - a review.

    Science.gov (United States)

    Woo, E H C; White, P; Lai, C W K

    2016-03-01

    This paper presents an overview of global ergonomics standards and guidelines for design of computer workstations, with particular focus on their inconsistency and associated health risk impact. Overall, considerable disagreements were found in the design specifications of computer workstations globally, particularly in relation to the results from previous ergonomics research and the outcomes from current ergonomics standards and guidelines. To cope with the rapid advancement in computer technology, this article provides justifications and suggestions for modifications in the current ergonomics standards and guidelines for the design of computer workstations. Practitioner Summary: A research gap exists in ergonomics standards and guidelines for computer workstations. We explore the validity and generalisability of ergonomics recommendations by comparing previous ergonomics research through to recommendations and outcomes from current ergonomics standards and guidelines.

  10. Improved safety of the system 80+TM standard plants design through increased diversity and redundancy of safety systems

    International Nuclear Information System (INIS)

    Matzie, Regis A.; Carpentino, Frederick L.; Robertson, James E.

    1996-01-01

    Safely systems in the System 80+ TM Standard Plant are designed with more redundancy, diversity and simplicity than earlier nuclear power plant designs. These gains were accomplished by an evolutionary process that preserved the desirable and proven features in currently operating nuclear plants, while improving reliability and defense-in-depth. The System 80+ safety systems are the primary contributors to a core damage frequency that is more than 100 times lower than 1980's vintage U. S. designs, including the predecessor System 80 R standard nuclear steam supply system (NSSS) design. The System 80+ design includes significant improvements to the safety injection system, emergency feedwater system, shutdown cooling system, containment spray system, reactor coolant gas vent system, and to their vital support systems. These improvements enhance performance for traditional design basis events and significantly reduce the probability of a severe accident. The System 80+ design also incorporates safety systems to mitigate a severe accident. The added systems include the rapid depressurization system, the in-containment refueling water storage tank, the cavity flooding system. These systems fully address the U. S. Nuclear Regulatory Commission's (US NRC) severe accident policy. The System 80+ safety systems are integrated with the System 80+ Nuclear Island (NI) design. The NI general arrangement provides quadrant separation of the safety systems for protection from fire and flooding, and large equipment pull spaces and lay down areas for maintenance. This paper will describe the System 80+ safety systems advanced design features, the improved accident prevention and mitigation capabilities, and startup, operating and maintenance benefits

  11. Methods for formulation of design basis accidents within a risk-informed approach to safety regulation of new nuclear power plants

    International Nuclear Information System (INIS)

    Beer, B.C.; Apostolakis, G.E.; Golay, M.W.

    2000-01-01

    Within a project sponsored by the U.S. Department of Energy (DOE) an investigation is being conducted into creating a risk-informed safety regulatory framework and design process based upon the use of probabilistic risk assessment (PRA). In conjunction with efforts to formulate an overall regulatory framework (i.e., reported in PSAM 5 by F. Duran, A. Camp, G. Apostolakis and M. Golay, 'A Framework for Regulatory Requirements and Industry Standards for New Nuclear Power Plants'), this paper addresses the potential role(s) of Design Basis Accidents (DBAs) within this new framework. Currently that role, if any, is unclear. In previous nuclear safety regulatory treatments, DBAs have been of great practical value for both designers and regulators. However, they have suffered from being inconsistently formulated, and lacking fundamental justification. Any DBA set is likely to be formulated uniquely for a specific reactor concept. The staff of any nuclear power plant (NPP) in the U.S. routinely calculates the likelihood of core damage, the likelihood of radioactive release and the likelihood of adverse health effects due to radioactive release. As the accuracy of such estimates improves industry-wide, safety regulators consider weighing these calculated risks more heavily than strict adherence to the prescriptive conservatisms of existing regulations, hence risk-informed regulation. DBAs, despite their prescriptive nature, can remain useful tools for regulators and designers in a risk-informed regulatory framework, providing that they can be formulated in a fashion consistent with the risk profiles of a plant. DBAs also offer the opportunity to take into account factors of uncertainty not captured in a PRA, which are typically addressed via defense-in-depth features and subjective judgements. Designers seeking only to create a plant having a calculated risk below a certain value, while minimizing cost, may find themselves in an inefficient trial-and-error process as they

  12. Modified Standard Penetration Test–based Drilled Shaft Design Method for Weak Rocks (Phase 2 Study)

    Science.gov (United States)

    2017-12-15

    In this project, Illinois-specific design procedures were developed for drilled shafts founded in weak shale or rock. In particular, a modified standard penetration test was developed and verified to characterize the in situ condition of weak shales ...

  13. HSTL IO Standard Based Energy Efficient Multiplier Design using Nikhilam Navatashcaramam Dashatah on 28nm FPGA

    DEFF Research Database (Denmark)

    Madhok, Shivani; Pandey, Bishwajeet; Kaur, Amanpreet

    2015-01-01

    standards. Frequency scaling is one of the best energy efficient techniques for FPGA based VLSI design and is used in this paper. At the end we can conclude that we can conclude that there is 23-40% saving of total power dissipation by using SSTL IO standard at 25 degree Celsius. The main reason for power...... consumption is leakage power at different IO Standards and at different frequencies. In this research work only FPGA work has been performed not ultra scale FPGA....

  14. 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.

  15. [Design and implementation of medical instrument standard information retrieval system based on APS.NET].

    Science.gov (United States)

    Yu, Kaijun

    2010-07-01

    This paper Analys the design goals of Medical Instrumentation standard information retrieval system. Based on the B /S structure,we established a medical instrumentation standard retrieval system with ASP.NET C # programming language, IIS f Web server, SQL Server 2000 database, in the. NET environment. The paper also Introduces the system structure, retrieval system modules, system development environment and detailed design of the system.

  16. The design of a calorimetric standard of ionising radiation absorbed dose

    International Nuclear Information System (INIS)

    Huntley, R.B.

    1981-05-01

    The design of a calorimetric working standard of ionising radiation absorbed dose is discussed. A brief history of the appropriate quantities and units of measurement is given. Detailed design considerations follow a summary of the relevant literature. The methods to be used to relate results to national standards of measurement are indicated, including the need for various correction factors. A status report is given on the construction and testing program

  17. Bias Corrections for Standardized Effect Size Estimates Used with Single-Subject Experimental Designs

    Science.gov (United States)

    Ugille, Maaike; Moeyaert, Mariola; Beretvas, S. Natasha; Ferron, John M.; Van den Noortgate, Wim

    2014-01-01

    A multilevel meta-analysis can combine the results of several single-subject experimental design studies. However, the estimated effects are biased if the effect sizes are standardized and the number of measurement occasions is small. In this study, the authors investigated 4 approaches to correct for this bias. First, the standardized effect…

  18. 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

  19. 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)

  20. 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.

  1. 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

  2. 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.

  3. Standardization of Schwarz-Christoffel transformation for engineering design of semiconductor and hybrid integrated-circuit elements

    Science.gov (United States)

    Yashin, A. A.

    1985-04-01

    A semiconductor or hybrid structure into a calculable two-dimensional region mapped by the Schwarz-Christoffel transformation and a universal algorithm can be constructed on the basis of Maxwell's electro-magnetic-thermal similarity principle for engineering design of integrated-circuit elements. The design procedure involves conformal mapping of the original region into a polygon and then the latter into a rectangle with uniform field distribution, where conductances and capacitances are calculated, using tabulated standard mapping functions. Subsequent synthesis of a device requires inverse conformal mapping. Devices adaptable as integrated-circuit elements are high-resistance film resistors with periodic serration, distributed-resistance film attenuators with high transformation ratio, coplanar microstrip lines, bipolar transistors, directional couplers with distributed coupling to microstrip lines for microwave bulk devices, and quasirregular smooth matching transitions from asymmetric to coplanar microstrip lines.

  4. A brief comparison of existing regional green building design standards in China

    Science.gov (United States)

    Wang, J.; Liu, Y.; Ren, J.; Cho, S.

    2017-03-01

    A large country with a variety of regional natural, cultural and economic conditions, China has established a number of green building design (GBD) standards both at national and regional (provincial and municipal) levels. Some researches have been conducted to review and compare such standards. The main aim was to provide valuable references for the establishment of new regionally specific GBD standards in different regions of the country. This paper introduces the preliminary results of the researches. The distribution, frameworks and content of the existing regional GBD standards are introduced and compared in relating to the regionally specific climate, resource, economic and cultural conditions. Conclusions are provided and further researches are recommended.

  5. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Science.gov (United States)

    2013-09-30

    ... Design Certification and Operating Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... Design Certification and Operating Reactors.'' The NRC seeks comments on the proposed revised section of... subject): Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013...

  6. Design analysis and comparison between standard and rotary porting systems for IC engine

    Energy Technology Data Exchange (ETDEWEB)

    Palmisano, R.; Ng, H.D. [Concordia Univ., Montreal, PQ (Canada). Dept. of Mechanical and Industrial Engineering

    2009-07-01

    A method of improving the efficiency of an internal combustion engine was presented. The proposed design used a new sealing technology to optimize a rotary valve design. The design was compared with standard poppet porting system using a computer aided engineering (CAE) method. A double port cylinder configuration was used in order to allow continuous flow into the chamber and minimize the exposed surface area in the combustion chamber. The design also minimized distortion on the component when exposed to high condensation pressures. A transfer case was used to allow for easy adaptation of the configuration to a multi-cylinder engine. A mechanism dynamics extension (MDO) program was used to conduct dynamic analyses on the rotary and standard valve porting systems. The study showed that the poppet system required 62 times more torque than the new rotary valve design. It was concluded that further research is needed to examine the flow properties of both designs. 2 refs., 6 figs.

  7. The applicability of ALPHA/PHOENIX/ANC nuclear design code system on Korean standard PWR's

    International Nuclear Information System (INIS)

    Lee, Kookjong; Choi, Kie-Yong; Lee, Hae-Chan; Roh, Eun-Rae

    1996-01-01

    For the Korean Standard Nuclear Power Plant (KSNPP) designed based on Combustion Engineering (CE) System 80, the Westinghouse nuclear design code system ALPHA/PHOENIX/ANC was applied to the follow-up design of initial and reload core of KSNPP. The follow-up design results of Yonggwang Unit 3 Cycle 1, 2 and Yonggwang Unit 4 Cycle 1 have shown good agreements with the measured data. The assemblywise power distributions have shown less than 2% average differences and critical boron concentrations have shown less than 20 ppm differences. All the low power physics test parameters are in good agreement. Consequently, APA design code system can be applied to KNSPP cores. (author)

  8. Index of Non-Government Standards on Human Engineering Design Criteria and Program Requirements/Guidelines. Version 3

    National Research Council Canada - National Science Library

    Poston, Alan

    2002-01-01

    .... Since the designation of documents as standards by non-government standards bodies tends to be somewhat flexible, the scope of non-government standards for the Index was kept quite loose and includes...

  9. The impact of safety standards updating for design purposes in nuclear power plants licensing

    International Nuclear Information System (INIS)

    Alvarenga, Marco Antonio Bayout; Rabello, Sidney Luiz

    2009-01-01

    The Brazilian experience of nuclear power plants licensing was consolidated by the use of the Brazilian, American, German and IAEA standards. Independently of the set of norms, standards or guides to be used, this set should be in consonance with the state-of-art or the current state of knowledge in science and technology. In the general design criteria of US NRC or German BMI, or in the Brazilian norms (CNEN) or even, in the IAEA standards, this aspect is always emphasized. On the other hand, the international operational experience of nuclear reactors (for example, TMI accident) also contributes to the updating of norms and standards. The use of new technologies (for example, digital technology) impels the norms and standards to adopt new design criteria related to the new technological context. Moreover, we must add the particular vision that each country can have concerning to specific topics in nuclear safety. This work discusses how the norms, standards and guides used in the nuclear licensing are being reviewed to cope with the requirement of the state-of-art. In order to accomplish this aim we took some general design criteria to exemplify how they are fulfilled, mainly those related directly with the protection of the defense-in-depth barriers: primary coolant system, containment vessel and containment systems, including external events and severe accidents. In complement to the deterministic analysis, it is also discussed the design criteria related to the human factors engineering and probabilistic safety analysis, including severe accidents aspects. (author)

  10. Recent developments in the IAEA safety standards: design and operation of nuclear power plants

    International Nuclear Information System (INIS)

    Saito, Takehiko

    2004-01-01

    The IAEA has been publishing a wide variety of safety standards for nuclear and radiation related facilities and activities since 1978. In 1996, a more rigorously structured approach for the preparation and review of its safety standards was introduced. Currently, based on the approach, revision of most of the standards is in completion or near completion. The latest versions of the Safety Requirements for ''Design'' and ''Operation'' of nuclear power plants were respectively published in 2000. Currently, along with this revision of the Safety Requirements, many Safety Guides have been revised. In order to clarify the complicated revision procedure, an example of the entire revision process for a Safety Guide is provided. Through actual example of the revision process, enormous amount of work involved in the revision work is clearly indicated. The current status of all of the Safety Standards for Design and that for Operation of nuclear power plants are summarized. Summary of other IAEA safety standards currently revised and available related IAEA publications, together with information on the IAEA Web Site from where these documents can be downloaded, is also provided. The standards are reviewed to determine whether revision (or new issue) is necessary in five years following publication. The IAEA safety standards will continue to be updated through comprehensive and structured approach, collaboration of many experts of the world, and reflecting good practices of the world. The IAEA safety standards will serve to provide high level of safety assurance. (author)

  11. The design of reliability data bases, part I: review of standard design concepts

    International Nuclear Information System (INIS)

    Cooke, Roger M.

    1996-01-01

    Main styles in the design of reliability data banks (RDB's) are reviewed. The conceptual and mathematical tools underlying these designs are summarized. A key point is the method for assessing failure rates for competing failure modes. The theory of independent competing risk and the relation to colored Poisson processes is explained. The notions of observed and naked failure rates are defined, and their equivalence under the assumption of independence is shown. In conclusion, the needs of different users are compared with the information currently offered

  12. Design validation of the ITER EC upper launcher according to codes and standards

    Energy Technology Data Exchange (ETDEWEB)

    Spaeh, Peter, E-mail: peter.spaeh@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Gagliardi, Mario [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); F4E, Fusion for Energy, Joint Undertaking, Barcelona (Spain); Grossetti, Giovanni; Meier, Andreas; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Weinhorst, Bastian [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2015-10-15

    Highlights: • A set of applicable codes and standards has been chosen for the ITER EC upper launcher. • For a particular component load combinations, failure modes and stress categorizations have been determined. • The design validation was performed in accordance with the “design by analysis”-approach of the ASME boiler and pressure vessel code section III. - Abstract: The ITER electron cyclotron (EC) upper launcher has passed the CDR (conceptual design review) in 2005 and the PDR (preliminary design review) in 2009 and is in its final design phase now. The final design will be elaborated by the European consortium ECHUL-CA with contributions from several research institutes in Germany, Italy, the Netherlands and Switzerland. Within this consortium KIT is responsible for the design of the structural components (the upper port plug, UPP) and also the design integration of the launcher. As the selection of applicable codes and standards was under discussion for the past decade, the conceptual and the preliminary design of the launcher structure were not elaborated in straight accordance with a particular code but with a variety of well-acknowledged engineering practices. For the final design it is compulsory to validate the design with respect to a typical engineering code in order to be compliant with the ITER quality and nuclear requirements and to get acceptance from the French regulator. This paper presents typical design validation of the closure plate, which is the vacuum and Tritium barrier and thus a safety relevant component of the upper port plug (UPP), performed with the ASME boiler and pressure vessel code. Rationales for choosing this code are given as well as a comparison between different design methods, like the “design by rule” and the “design by analysis” approach. Also the selections of proper load specifications and the identification of potential failure modes are covered. In addition to that stress categorizations, analyses

  13. Design validation of the ITER EC upper launcher according to codes and standards

    International Nuclear Information System (INIS)

    Spaeh, Peter; Aiello, Gaetano; Gagliardi, Mario; Grossetti, Giovanni; Meier, Andreas; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro; Weinhorst, Bastian

    2015-01-01

    Highlights: • A set of applicable codes and standards has been chosen for the ITER EC upper launcher. • For a particular component load combinations, failure modes and stress categorizations have been determined. • The design validation was performed in accordance with the “design by analysis”-approach of the ASME boiler and pressure vessel code section III. - Abstract: The ITER electron cyclotron (EC) upper launcher has passed the CDR (conceptual design review) in 2005 and the PDR (preliminary design review) in 2009 and is in its final design phase now. The final design will be elaborated by the European consortium ECHUL-CA with contributions from several research institutes in Germany, Italy, the Netherlands and Switzerland. Within this consortium KIT is responsible for the design of the structural components (the upper port plug, UPP) and also the design integration of the launcher. As the selection of applicable codes and standards was under discussion for the past decade, the conceptual and the preliminary design of the launcher structure were not elaborated in straight accordance with a particular code but with a variety of well-acknowledged engineering practices. For the final design it is compulsory to validate the design with respect to a typical engineering code in order to be compliant with the ITER quality and nuclear requirements and to get acceptance from the French regulator. This paper presents typical design validation of the closure plate, which is the vacuum and Tritium barrier and thus a safety relevant component of the upper port plug (UPP), performed with the ASME boiler and pressure vessel code. Rationales for choosing this code are given as well as a comparison between different design methods, like the “design by rule” and the “design by analysis” approach. Also the selections of proper load specifications and the identification of potential failure modes are covered. In addition to that stress categorizations, analyses

  14. A Standardized Rubric for Evaluating Webquest Design: Reliability Analysis of ZUNAL Webquest Design Rubric

    Science.gov (United States)

    Unal, Zafer; Bodur, Yasar; Unal, Aslihan

    2012-01-01

    Current literature provides many examples of rubrics that are used to evaluate the quality of web-quest designs. However, reliability of these rubrics has not yet been researched. This is the first study to fully characterize and assess the reliability of a webquest evaluation rubric. The ZUNAL rubric was created to utilize the strengths of the…

  15. Establishment of quality, reliability and design standards for low, medium, and high power microwave hybrid microcircuits

    Science.gov (United States)

    Robinson, E. A.

    1973-01-01

    Quality, reliability, and design standards for microwave hybrid microcircuits were established. The MSFC Standard 85M03926 for hybrid microcircuits was reviewed and modifications were generated for use with microwave hybrid microcircuits. The results for reliability tests of microwave thin film capacitors, transistors, and microwave circuits are presented. Twenty-two microwave receivers were tested for 13,500 unit hours. The result of 111,121 module burn-in and operating hours for an integrated solid state transceiver module is reported.

  16. Standardization Efforts for Mechanical Testing and Design of Advanced Ceramic Materials and Components

    Science.gov (United States)

    Salem, Jonathan A.; Jenkins, Michael G.

    2003-01-01

    Advanced aerospace systems occasionally require the use of very brittle materials such as sapphire and ultra-high temperature ceramics. Although great progress has been made in the development of methods and standards for machining, testing and design of component from these materials, additional development and dissemination of standard practices is needed. ASTM Committee C28 on Advanced Ceramics and ISO TC 206 have taken a lead role in the standardization of testing for ceramics, and recent efforts and needs in standards development by Committee C28 on Advanced Ceramics will be summarized. In some cases, the engineers, etc. involved are unaware of the latest developments, and traditional approaches applicable to other material systems are applied. Two examples of flight hardware failures that might have been prevented via education and standardization will be presented.

  17. Design and evaluation of a 10-mA DC current reference standard

    CERN Document Server

    Fernqvist, G; Pickering, J; Power, F

    2003-01-01

    A new DC current reference standard has been developed for high- current power converter calibration in the large hadron collider (LHC) project at the European Organization for Nuclear Research (CERN). This standard provides a near ideal 10-mA DC current with long-term drift of one part in 10/sup 6/ per year. The paper describes the requirements and the detailed design and evaluation of the unit. Since similar 10-V standards are commercially available, the paper concentrates on the unique current output capability of this device. (4 refs).

  18. Report on design rules of μ-tools for standard insert

    DEFF Research Database (Denmark)

    Tosello, Guido; Esmoris, Josa Ignacio; Quadroni, A.

    2011-01-01

    -effectively, especially for micro injection moulding. This particular deliverable has the objective to present the design rules for high performance μ-tools and inserts manufacture based on the new standard manufacturing process chains established during the WP 2.2 work. In particular, the achievable features, surfaces......Tooling is one of the critical stages of the process chain for polymer micro products manufacture and in particular for the COTECH process chain. Therefore, within the scope of SP2 “Tooling”, the WP 2.2 “New tool-making solutions for μ-IM and HE” is designed to investigate, develop and standardize...

  19. SSTL I/O Standard Based Green Communication Using Fibonacci Generator Design on Ultra Scale FPGA

    DEFF Research Database (Denmark)

    Nagah, Sumi; Kaur, Ravinder; Pandey, Bishwajeet

    2015-01-01

    In this paper six different available classes of Stub-Series Terminated Logic (SSTL) Input/output standard is used for the design of Green Fibonacci generator on 40nm FGPA. That green Fibonacci Generator is used to generate key for Wi-Fi Protected Access in order to make energy efficient communic......In this paper six different available classes of Stub-Series Terminated Logic (SSTL) Input/output standard is used for the design of Green Fibonacci generator on 40nm FGPA. That green Fibonacci Generator is used to generate key for Wi-Fi Protected Access in order to make energy efficient...

  20. Review of application code and standards for mechanical and piping design of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the propriety of application code and standards for mechanical and piping of HANARO fuel test loop and to decide the technical specification of FTL systems. (author). 18 refs., 8 tabs., 6 figs.

  1. How the new optoelectronic design automation industry is taking advantage of preexisting EDA standards

    Science.gov (United States)

    Nesmith, Kevin A.; Carver, Susan

    2014-05-01

    With the advancements in design processes down to the sub 7nm levels, the Electronic Design Automation industry appears to be coming to an end of advancements, as the size of the silicon atom becomes the limiting factor. Or is it? The commercial viability of mass-producing silicon photonics is bringing about the Optoelectronic Design Automation (OEDA) industry. With the science of photonics in its infancy, adding these circuits to ever-increasing complex electronic designs, will allow for new generations of advancements. Learning from the past 50 years of the EDA industry's mistakes and missed opportunities, the photonics industry is starting with electronic standards and extending them to become photonically aware. Adapting the use of pre-existing standards into this relatively new industry will allow for easier integration into the present infrastructure and faster time to market.

  2. Digitalization as Driver for Standardized Specification and Design of Buildings: In Search of an Efficient Building Design Management Methodology

    DEFF Research Database (Denmark)

    Treldal, Niels

    of this research is, therefore, to increase the understanding of the relation between information needs, standardisation and efficient design management. The research draws on findings from previous research on information management, design management and socio-technical science and focuses in particular......-value adding design iterations will occur too frequently if the variability is not carefully managed. Building a strong community within the design team is found to be critical to reduce variability as it allows project managers to entrust the team to find solutions and coordinate activities more efficiently...... standards were developed in the current research. An IDM package framework is proposed to make the current IDM methodology from buildingSMART more modular and easier to reuse and utilize on projects. A generic LOD framework is proposed to make the agreement on geometry information exchange more pragmatic...

  3. Contributions of the ORNL piping program to nuclear piping design codes and standards

    International Nuclear Information System (INIS)

    Moore, S.E.

    1975-11-01

    The ORNL Piping Program was conceived and established to develop basic information on the structural behavior of nuclear power plant piping components and to prepare this information in forms suitable for use in design analysis and codes and standards. One of the objectives was to develop and qualify stress indices and flexibility factors for direct use in Code-prescribed design analysis methods. Progress in this area is described

  4. Applicability of ISO 9001: 2000 standard to design and research activities in nuclear objectives and installations

    International Nuclear Information System (INIS)

    Stamatopol, M.; Patticu, C.

    2001-01-01

    The ISO 9001: 2000 standard contains the results of the latest studies concerning the design and implementation of quality management systems. The continuous improvement of these systems required a new approach, namely, the process-based approach. A process-based model of quality management is presented in terms of client request, input data, management responsibility, resource management, product realisation, measurement / analysis and optimization, product, output data and client request fulfilment. Thus, the quality management system becomes an ensemble of interconnected or interacting processes. Consequently, to implement such a system the necessary processes have to be identified and their ISO 9001: 2000 standard based management ensured. Process based approach also allows better evidencing the input and output data specific to each process. Such an approach grants a wider applicability to the design and research activities in nuclear objectives and installations. This work aims at identifying additional requirements implied in design and research activities in nuclear field, as stipulated in the following standards / documents: - IAEA SG Q8 'Quality assurance in research and development'; - IAEA SG Q10 'Quality assurance in design'; - CAN / CSA N 286.2 - 00 'Design Quality Assurance for Nuclear Power Plants'; CAN 3 - CSA Z 299.1 'Quality Assurance Program - Category 1'; - NQA - 1, SUPPLEMENT 3S-1 'Supplementary requirements for design control'

  5. Research and development issues for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Ando, Masanori; Morishita, Masaki

    2003-01-01

    For realization of safe and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts and validate their structural integrity, the research and development of 'Fast Reactor Structural Design Standard (FDS)' is recognized as an essential theme. FDS considers general characteristics of FRs and design needs for their rationalization. Three main subjects were settled in research and development issues of FDS. One is rationalization of failure criteria' taking characteristic design conditions into account. Next is development of 'a guideline on inelastic analysis for design' in order to predict elastic plastic and creep behaviours of high temperature components. Furthermore, efforts are being made toward preparing a guideline on thermal loads modeling' for FR component design where thermal loads are dominant. (author)

  6. Design and initial characterization of the SC-200 proteomics standard mixture.

    Science.gov (United States)

    Bauman, Andrew; Higdon, Roger; Rapson, Sean; Loiue, Brenton; Hogan, Jason; Stacy, Robin; Napuli, Alberto; Guo, Wenjin; van Voorhis, Wesley; Roach, Jared; Lu, Vincent; Landorf, Elizabeth; Stewart, Elizabeth; Kolker, Natali; Collart, Frank; Myler, Peter; van Belle, Gerald; Kolker, Eugene

    2011-01-01

    High-throughput (HTP) proteomics studies generate large amounts of data. Interpretation of these data requires effective approaches to distinguish noise from biological signal, particularly as instrument and computational capacity increase and studies become more complex. Resolving this issue requires validated and reproducible methods and models, which in turn requires complex experimental and computational standards. The absence of appropriate standards and data sets for validating experimental and computational workflows hinders the development of HTP proteomics methods. Most protein standards are simple mixtures of proteins or peptides, or undercharacterized reference standards in which the identity and concentration of the constituent proteins is unknown. The Seattle Children's 200 (SC-200) proposed proteomics standard mixture is the next step toward developing realistic, fully characterized HTP proteomics standards. The SC-200 exhibits a unique modular design to extend its functionality, and consists of 200 proteins of known identities and molar concentrations from 6 microbial genomes, distributed into 10 molar concentration tiers spanning a 1,000-fold range. We describe the SC-200's design, potential uses, and initial characterization. We identified 84% of SC-200 proteins with an LTQ-Orbitrap and 65% with an LTQ-Velos (false discovery rate = 1% for both). There were obvious trends in success rate, sequence coverage, and spectral counts with protein concentration; however, protein identification, sequence coverage, and spectral counts vary greatly within concentration levels.

  7. Safety evaluation report related to the preliminary design of the Standard Reference System, RESAR-414

    International Nuclear Information System (INIS)

    1978-11-01

    The safety evaluation for the Westinghouse Standard Reactor includes information on general reactor characteristics; design criteria for systems and components; reactor coolant system; engineered safety systems; instrumentation and controls; electric power systems; auxiliary systems; steam and power conversion system; radioactive waste management; radiation protection; conduct of operations; accident analyses; and quality assurance

  8. Study on Design and Implementation of JAVA Programming Procedural Assessment Standard

    Science.gov (United States)

    Tingting, Xu; Hua, Ma; Xiujuan, Wang; Jing, Wang

    2015-01-01

    The traditional JAVA course examination is just a list of questions from which we cannot know students' skills of programming. According to the eight abilities in curriculum objectives, we designed an assessment standard of JAVA programming course that is based on employment orientation and apply it to practical teaching to check the teaching…

  9. 75 FR 34657 - Energy Efficiency and Sustainable Design Standards for New Federal Buildings

    Science.gov (United States)

    2010-06-18

    ... Efficiency and Sustainable Design Standards for New Federal Buildings AGENCY: Office of Energy Efficiency and....S. Department of Energy, Office of Energy Efficiency and Renewable Energy, Federal Energy Management... June 11, 2010. Cathy Zoi, Assistant Secretary, Energy Efficiency and Renewable Energy. [FR Doc. 2010...

  10. A Novel Instructional Approach to the Design of Standard Controllers: Using Inversion Formulae

    Science.gov (United States)

    Ntogramatzidis, Lorenzo; Zanasi, Roberto; Cuoghi, Stefania

    2014-01-01

    This paper describes a range of design techniques for standard compensators (Lead-Lag networks and PID controllers) that have been applied to the teaching of many undergraduate control courses throughout Italy over the last twenty years, but that have received little attention elsewhere. These techniques hinge upon a set of simple formulas--herein…

  11. 76 FR 76048 - Air Quality Designations for the 2008 Lead (Pb) National Ambient Air Quality Standards

    Science.gov (United States)

    2011-12-06

    ... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Part 81 [EPA-HQ-OAR-2009-0443; FRL-9492-3] RIN 2060-AR17 Air Quality Designations for the 2008 Lead (Pb) National Ambient Air Quality Standards Correction In rule document 2011-29460 appearing on pages 72097-72120 in the issues of Tuesday, November 22, 2011...

  12. A Discussion of Change Theory, System Theory, and State Designed Standards and Accountability Initiatives.

    Science.gov (United States)

    McNeal, Larry; Christy, W. Keith

    This brief paper is a presentation that preceeded another case of considering the ongoing dialogue on the advantages and disadvantages of centralized and decentralized school-improvement processes. It attempts to raise a number of questions about the relationship between state-designed standards and accountability initiatives and change and…

  13. 78 FR 7251 - Rulemaking Concerning the Standards for Designating Positions in the Competitive Service as...

    Science.gov (United States)

    2013-01-31

    ...--Rulemaking Concerning the Standards for Designating Positions in the Competitive Service as National Security... Positions in the Competitive Service as National Security Sensitive and Related Matters Memorandum for the Director of National Intelligence [and] the Director of the Office of Personnel Management The Director of...

  14. A pro-forma design for car-carriers: Low-speed performance-based standards

    CSIR Research Space (South Africa)

    Benade R, Berman R

    2015-07-01

    Full Text Available the constraints of the pro-forma design met the Level 1 requirements of the low-speed PBS. Future work will ensure compliance with the full set of twelve performance standards. It is estimated that the pro-forma approach as compared to doing full assessments would...

  15. Design and Implementation of a Prototype with a Standardized Interface for Transducers in Ambient Assisted Living

    Directory of Open Access Journals (Sweden)

    Enrique Dorronzoro

    2015-01-01

    Full Text Available Solutions in the field of Ambient Assisted Living (AAL do not generally use standards to implement a communication interface between sensors and actuators. This makes these applications isolated solutions because it is so difficult to integrate them into new or existing systems. The objective of this research was to design and implement a prototype with a standardized interface for sensors and actuators to facilitate the integration of different solutions in the field of AAL. Our work is based on the roadmap defined by AALIANCE, using motes with TinyOS telosb, 6LoWPAN, sensors, and the IEEE 21451 standard protocol. This prototype allows one to upgrade sensors to a smart status for easy integration with new applications and already existing ones. The prototype has been evaluated for autonomy and performance. As a use case, the prototype has been tested in a serious game previously designed for people with mobility problems, and its advantages and disadvantages have been analysed.

  16. Towards product design automation based on parameterized standard model with diversiform knowledge

    Science.gov (United States)

    Liu, Wei; Zhang, Xiaobing

    2017-04-01

    Product standardization based on CAD software is an effective way to improve design efficiency. In the past, research and development on standardization mainly focused on the level of component, and the standardization of the entire product as a whole is rarely taken into consideration. In this paper, the size and structure of 3D product models are both driven by the Excel datasheets, based on which a parameterized model library is therefore established. Diversiform knowledge including associated parameters and default properties are embedded into the templates in advance to simplify their reuse. Through the simple operation, we can obtain the correct product with the finished 3D models including single parts or complex assemblies. Two examples are illustrated later to invalid the idea, which will greatly improve the design efficiency.

  17. Analysis of the Current Technical Issues on ASME Code and Standard for Nuclear Mechanical Design(2009)

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2009-11-01

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  18. Analysis of the Current Technical Issues on ASME Code and Standard for Nuclear Mechanical Design(2009)

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2009-11-15

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  19. 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

  20. Standard concerning the design of nuclear power stations in earthquake-prone districts

    International Nuclear Information System (INIS)

    Kirillov, A.P.; Anbriashivili, Y.K.; Suvilova, A.V.

    1980-01-01

    The measures of security assurance against the effect of radioactive contamination has become more and more complex due to the construction of nuclear power stations of diverse types. The aseismatic measures for the nuclear power stations built in the districts where earthquakes of different intensity occur are important problems. All main machinery and equipments and emergency systems in power stations must be protected from earthquakes, and this makes the solution of problems difficult. At present in USSR, the provisional standard concerning the design of atomic energy facilities built in earthquake-prone districts is completed. The basic philosophy of the standard is to decide the general requirements as the conditions for the design of nuclear power stations built in earthquake-prone districts. The lowest earthquake activity in the construction districts is considered as magnitude 4, and in the districts where earthquake activity is magnitude 9 or more, the construction of nuclear power stations is prohibited. Two levels of earthquake action are specified for the design: design earthquake and the largest design earthquake. The construction sites of nuclear power stations must be 15 to 150 km distant from the potential sources of earthquakes. Nuclear power stations are regarded as the aseismatically guaranteed type when the safety of reactors is secured under the application of the standard. The buildings and installations are classified into three classes regarding the aseismatic properties. (Kako, I.)

  1. 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

  2. The Synthetic Biology Open Language (SBOL) provides a community standard for communicating designs in synthetic biology.

    Science.gov (United States)

    Galdzicki, Michal; Clancy, Kevin P; Oberortner, Ernst; Pocock, Matthew; Quinn, Jacqueline Y; Rodriguez, Cesar A; Roehner, Nicholas; Wilson, Mandy L; Adam, Laura; Anderson, J Christopher; Bartley, Bryan A; Beal, Jacob; Chandran, Deepak; Chen, Joanna; Densmore, Douglas; Endy, Drew; Grünberg, Raik; Hallinan, Jennifer; Hillson, Nathan J; Johnson, Jeffrey D; Kuchinsky, Allan; Lux, Matthew; Misirli, Goksel; Peccoud, Jean; Plahar, Hector A; Sirin, Evren; Stan, Guy-Bart; Villalobos, Alan; Wipat, Anil; Gennari, John H; Myers, Chris J; Sauro, Herbert M

    2014-06-01

    The re-use of previously validated designs is critical to the evolution of synthetic biology from a research discipline to an engineering practice. Here we describe the Synthetic Biology Open Language (SBOL), a proposed data standard for exchanging designs within the synthetic biology community. SBOL represents synthetic biology designs in a community-driven, formalized format for exchange between software tools, research groups and commercial service providers. The SBOL Developers Group has implemented SBOL as an XML/RDF serialization and provides software libraries and specification documentation to help developers implement SBOL in their own software. We describe early successes, including a demonstration of the utility of SBOL for information exchange between several different software tools and repositories from both academic and industrial partners. As a community-driven standard, SBOL will be updated as synthetic biology evolves to provide specific capabilities for different aspects of the synthetic biology workflow.

  3. 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

  4. Meeting Classroom Needs: Designing Space Physics Educational Outreach for Science Education Standards

    Science.gov (United States)

    Urquhart, M. L.; Hairston, M.

    2008-12-01

    As with all NASA missions, the Coupled Ion Neutral Dynamics Investigation (CINDI) is required to have an education and public outreach program (E/PO). Through our partnership between the University of Texas at Dallas William B. Hanson Center for Space Sciences and Department of Science/Mathematics Education, the decision was made early on to design our educational outreach around the needs of teachers. In the era of high-stakes testing and No Child Left Behind, materials that do not meet the content and process standards teachers must teach cannot be expected to be integrated into classroom instruction. Science standards, both state and National, were the fundamental drivers behind the designs of our curricular materials, professional development opportunities for teachers, our target grade levels, and even our popular informal educational resource, the "Cindi in Space" comic book. The National Science Education Standards include much more than content standards, and our E/PO program was designed with this knowledge in mind as well. In our presentation we will describe how we came to our approach for CINDI E/PO, and how we have been successful in our efforts to have CINDI materials and key concepts make the transition into middle school classrooms. We will also present on our newest materials and high school physics students and professional development for their teachers.

  5. Development of the design of standardized units for the production of artificial radionuclides

    International Nuclear Information System (INIS)

    Auger, J.P.

    1976-01-01

    The production of artificial radionuclides began more than 20 years ago and has seen continuous growth at the rate over 20% a year. Technology has had to be adapted constantly to this growth in order to guarantee production and at the same time ensure the safety of personnel. The Department, which started its career in underground workings at Chatillon and then moved to the Saclay hot laboratories, is now housed in a building designed specially for the production of artificial radionuclides and equipped with standard production units. The first generation of standard units was sufficient to handle production which had begun to grow. Subsequently, thanks to the experience gained, there came into being a second generation of standardized units perfectly adapted to the new production requirements. The paper describes the evolution of design solutions between the first and the second standard, relating to contained cells, cell containment, remote control, interchangeability of cells, ventilation, waste discharge systems and repair of internal equipment. A highly positive evaluation can be made of the experience gained from the present standard. (author)

  6. Sandia National Laboratories Facilities Management and Operations Center Design Standards Manual

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Timothy L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-09-01

    At Sandia National Laboratories in New Mexico (SNL/NM), the design, construction, operation, and maintenance of facilities is guided by industry standards, a graded approach, and the systematic analysis of life cycle benefits received for costs incurred. The design of the physical plant must ensure that the facilities are "fit for use," and provide conditions that effectively, efficiently, and safely support current and future mission needs. In addition, SNL/NM applies sustainable design principles, using an integrated whole-building design approach, from site planning to facility design, construction, and operation to ensure building resource efficiency and the health and productivity of occupants. The safety and health of the workforce and the public, any possible effects on the environment, and compliance with building codes take precedence over project issues, such as performance, cost, and schedule. These design standards generally apply to all disciplines on all SNL/NM projects. Architectural and engineering design must be both functional and cost-effective. Facility design must be tailored to fit its intended function, while emphasizing low-maintenance, energy-efficient, and energy-conscious design. Design facilities that can be maintained easily, with readily accessible equipment areas, low maintenance, and quality systems. To promote an orderly and efficient appearance, architectural features of new facilities must complement and enhance the existing architecture at the site. As an Architectural and Engineering (A/E) professional, you must advise the Project Manager when this approach is prohibitively expensive. You are encouraged to use professional judgment and ingenuity to produce a coordinated interdisciplinary design that is cost-effective, easily contractible or buildable, high-performing, aesthetically pleasing, and compliant with applicable building codes. Close coordination and development of civil, landscape, structural, architectural, fire

  7. HCPB TBM thermo mechanical design: Assessment with respect codes and standards and DEMO relevancy

    International Nuclear Information System (INIS)

    Cismondi, F.; Kecskes, S.; Aiello, G.

    2011-01-01

    In the frame of the activities of the European TBM Consortium of Associates the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) is developed in Karlsruhe Institute of Technology (KIT). After performing detailed thermal and fluid dynamic analyses of the preliminary HCPB TBM design, the thermo mechanical behaviour of the TBM under typical ITER loads has to be assessed. A synthesis of the different design options proposed has been realized building two different assemblies of the HCPB-TBM: these two assemblies and the analyses performed on them are presented in this paper. Finite Element thermo-mechanical analyses of two detailed 1/4 scaled models of the HCPB-TBM assemblies proposed have been performed, with the aim of verifying the accordance of the mechanical behaviour with the criteria of the design codes and standards. The structural design limits specified in the codes and standard are discussed in relation with the EUROFER available data and possible damage modes. Solutions to improve the weak structural points of the present design are identified and the DEMO relevancy of the present thermal and structural design parameters is discussed.

  8. Innovating cystic fibrosis clinical trial designs in an era of successful standard of care therapies.

    Science.gov (United States)

    VanDevanter, Donald R; Mayer-Hamblett, Nicole

    2017-11-01

    Evolving cystic fibrosis 'standards of care' have influenced recent cystic fibrosis clinical trial designs for new therapies; care additions/improvements will require innovative trial designs to maximize feasibility and efficacy detection. Three cystic fibrosis therapeutic areas (pulmonary exacerbations, Pseudomonas aeruginosa airway infections, and reduced cystic fibrosis transmembrane conductance regulator [CFTR] protein function) differ with respect to the duration for which recognized 'standards of care' have been available. However, developers of new therapies in all the three areas are affected by similar challenges: standards of care have become so strongly entrenched that traditional placebo-controlled studies in cystic fibrosis populations likely to benefit from newer therapies have become less and less feasible. Today, patients/clinicians are more likely to entertain participation in active-comparator trial designs, that have substantial challenges of their own. Foremost among these are the selection of 'valid' active comparator(s), estimation of a comparator's current clinical efficacy (required for testing noninferiority hypotheses), and effective blinding of commercially available comparators. Recent and future cystic fibrosis clinical trial designs will have to creatively address this collateral result of successful past development of effective cystic fibrosis therapies: patients and clinicians are much less likely to accept simple, placebo-controlled studies to evaluate future therapies.

  9. 10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FACILITIES Pt. 50, App.N Appendix N to Part 50—Standardization of Nuclear Power Plant Designs: Permits To...

  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. Standards and the design of the Advanced Photon Source control system

    International Nuclear Information System (INIS)

    McDowell, W.P.; Knott, M.J.; Lenkszus, F.R.; Kraimer, M.R.; Daly, R.T.; Arnold, N.D.; Anderson, M.D.; Anderson, J.B.; Zieman, R.C.; Cha, Ben-Chin K.; Vong, F.C.; Nawrocki, G.J.; Gunderson, G.R.; Karonis, N.T.; Winans, J.R.

    1991-01-01

    The Advanced Photon Source (APS), now under construction at Argonne National Laboratory is a 7 GeV positron storage ring dedicated to research facilities using synchrotron radiation. This ring, along with its injection accelerators is to be controlled and monitored with a single, flexible, and expandable control system. In the conceptual stage the control system design group faced the challenges that face all control system designers: to force the machine designers to quantify and codify the system requirements, to protect the investment in hardware and software from rapid obsolescence, and to find methods of quickly incorporating new generations of equipment and replace of obsolete equipment without disrupting the exiting system. To solve these and related problems, the APS control system group made an early resolution to use standards in the design of the system. This paper will cover the present status of the APS control system as well as discuss the design decisions which led us to use industrial standards and collaborations with other laboratories whenever possible to develop a control system. It will explain the APS control system and illustrate how the use of standards has allowed APS to design a control system whose implementation addresses these issues. The system will use high performance graphic workstations using an X-Windows Graphical User Interface at the operator interface level. It connects to VME-based microprocessors at the field level using TCP/IP protocols over high performance networks. This strategy assures the flexibility and expansibility of the control system. A defined interface between the system components will allow the system to evolve with the direct addition of future, improved equipment and new capabilities

  12. 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

  13. 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.

  14. Design of a new torque standard machine based on a torque generation method using electromagnetic force

    International Nuclear Information System (INIS)

    Nishino, Atsuhiro; Ueda, Kazunaga; Fujii, Kenichi

    2017-01-01

    To allow the application of torque standards in various industries, we have been developing torque standard machines based on a lever deadweight system, i.e. a torque generation method using gravity. However, this method is not suitable for expanding the low end of the torque range, because of the limitations to the sizes of the weights and moment arms. In this study, the working principle of the torque generation method using an electromagnetic force was investigated by referring to watt balance experiments used for the redefinition of the kilogram. Applying this principle to a rotating coordinate system, an electromagnetic force type torque standard machine was designed and prototyped. It was experimentally demonstrated that SI-traceable torque could be generated by converting electrical power to mechanical power. Thus, for the first time, SI-traceable torque was successfully realized using a method other than that based on the force of gravity. (paper)

  15. Development of reliability-based design and assessment standards for onshore gas transmission pipelines

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Joe; Rothwell, Brian [TransCanada PipeLines Ltd., Calgary, AB (Canada); Nessim, Maher; Zhou, Wenxing [C-FER Technologies, Edmonton, AB (Canada)

    2005-07-01

    Onshore pipelines have traditionally been designed with a deterministic stress based methodology. The changing operating environment has however imposed many challenges to the pipeline industry, including heightened public awareness of risk, more challenging natural hazards and increased economic competitiveness. To meet the societal expectation of pipeline safety and enhance the competitiveness of the pipeline industry, significant efforts have been spent for the development of reliability-based design and assessment (RBDA) methodology. This paper will briefly review the technology development in the RBDA area and the focus will be on the progresses in the past years in standard development within the American Society of Mechanical Engineers (ASME) and the Canadian Standard Association (CSA) organizations. (author)

  16. SSTL I/O Standard Based Environment Friendly Energy Efficient ROM Design on FPGA

    DEFF Research Database (Denmark)

    Bansal, Neha; Bansal, Meenakshi; Saini, Rishita

    2014-01-01

    are operating ROM with the highest operating frequency of 4th generation i7 processor to test the compatibility of this design with the latest hardware in use. When there is no demand of peak performance, then we can save 74.5% clock power, 75% signal power, and 30.83% I/O power by operating our device with 1......Stub Series Terminated Logic (SSTL) is an Input/output standard. It is used to match the impedance of line, port and device of our design under consideration. Therefore, selection of energy efficient SSTL I/O standard among available different class of SSTL logic family in FPGA, plays a vital role...

  17. Standards and the design of the advanced photon source control system

    International Nuclear Information System (INIS)

    McDowell, W.P.; Knott, M.J.; Lenkszus, F.R.

    1992-01-01

    The Advanced Photon Source (APS), now under construction at Argonne National Laboratory (ANL), is a 7 GeV positron storage ring dedicated to research facilities using synchrotron radiation. This ring, along with its injection accelerators is to be controlled and monitored with a single, flexible, and expandable control system. This paper will cover the present status of the APS control system as well as discuss the design decisions which led us to use industrial standards and collaborations with other laboratories whenever possible to develop a control system. It will explain the APS control system and illustrate how the use of standards has allowed APS to design a control system whose implementation addresses these issues. (J.P.N.)

  18. DTU-ESA millimeter-wave validation standard antenna – requirements and design

    DEFF Research Database (Denmark)

    Pivnenko, Sergey; Kim, Oleksiy S.; Breinbjerg, Olav

    2014-01-01

    from a validation campaign is achieved when a dedicated Validation Standard (VAST) antenna specifically designed for this purpose is available. The driving requirements to VAST antennas are their mechanical stability with respect to any orientation of the antenna in the gravity field and thermal...... are briefly reviewed and the baseline design is described. The emphasis is given to definition of the requirements for the mechanical and thermal stability of the antenna, which satisfy the stringent stability requirement for the mm-VAST electrical characteristics....

  19. Impact of National Ambient Air Quality Standards Nonattainment Designations on Particulate Pollution and Health.

    Science.gov (United States)

    Zigler, Corwin M; Choirat, Christine; Dominici, Francesca

    2018-03-01

    Despite dramatic air quality improvement in the United States over the past decades, recent years have brought renewed scrutiny and uncertainty surrounding the effectiveness of specific regulatory programs for continuing to improve air quality and public health outcomes. We employ causal inference methods and a spatial hierarchical regression model to characterize the extent to which a designation of "nonattainment" with the 1997 National Ambient Air Quality Standard for ambient fine particulate matter (PM2.5) in 2005 causally affected ambient PM2.5 and health outcomes among over 10 million Medicare beneficiaries in the Eastern United States in 2009-2012. We found that, on average across all retained study locations, reductions in ambient PM2.5 and Medicare health outcomes could not be conclusively attributed to the nonattainment designations against the backdrop of other regional strategies that impacted the entire Eastern United States. A more targeted principal stratification analysis indicates substantial health impacts of the nonattainment designations among the subset of areas where the designations are estimated to have actually reduced ambient PM2.5 beyond levels achieved by regional measures, with noteworthy reductions in all-cause mortality, chronic obstructive pulmonary disorder, heart failure, ischemic heart disease, and respiratory tract infections. These findings provide targeted evidence of the effectiveness of local control measures after nonattainment designations for the 1997 PM2.5 air quality standard.

  20. 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