WorldWideScience

Sample records for safety-related components summary

  1. Evidence of aging effects on certain safety-related components: summary and analysis

    1995-09-01

    In response to interest shown by the Nuclear Energy Agency (NEA), Principal Working Group I (PWG- 1) of the Committee on the Safety of Nuclear Installations (CSNI) conducted a generic study on the effects of aging of active components in nuclear power plants. Representatives from France, Sweden, Finland, Japan, the United States, and the United Kingdom participated in the study by submitting reports documenting aging studies performed in their countries. This report consists of summaries of those reports, along with a comparison of the various statistical analysis methods used in the studies. The studies indicate that with some exceptions, active components generally do not present a significant aging problem in nuclear power plants. Design criteria and effective preventative maintenance programs, including timely replacement of components, are effective in mitigating potential aging problems. However, aging studies (such as qualitative and statistical analyses of failure modes and maintenance data) are an important part of efforts to identify and solve potential aging problems. Solving these problems typically includes such strategies as replacing suspect components with improved components, and implementing improved maintenance programs

  2. Experience on environmental qualification of safety-related components for Darlington Nuclear Generating Station

    Yu, A.S.; Kukreti, B.M.

    1987-01-01

    The proliferation of Nuclear Power Plant safety concerns has lead to increasing attention over the Environmental Qualification (EQ) of Nuclear Power Plant Safety-Related Components to provide the assurance that the safety related equipment will meet their intended functions during normal operation and postulated accident conditions. The environmental qualification of these components is also a Licensing requirement for Darlington Nuclear Generating Station. This paper provides an overview of EQ and the experience of a pilot project, in the qualification of the Main Moderator System safety-related functions for the Darlington Nuclear Generating Station currently under construction. It addresses the various phases of qualification from the identification of the EQ Safety-Related Components List, definition of location specific service conditions (normal, adbnormal and accident), safety-related functions, Environmental Qualification Assessments and finally, an EQ system summary report for the Main Moderator System. The results of the pilot project are discussed and the methodology reviewed. The paper concludes that the EQ Program developed for Darlington Nuclear Generating Station, as applied to the qualification of the Main Moderator System, contained all the elements necessary in the qualification of safety-related equipment. The approach taken in the qualification of the Moderator safety-related equipment proves to provide a sound framework for the qualification of other safety-related components in the station

  3. Passive components of NPP safety-related systems

    Ionaytis Romuald, R.; Bubnova Tatyana, A.

    2005-01-01

    This paper presents a new passive components with having drives: fast-response cutoff valves; modular actuators with opposite cocking pneumatic drives and actuation spring drives; voting electromagnetic valve units for control of pneumatic drives; passive initiators of actuation; visual diagnostics . All these devices have been developed and tested at mock-ups. This paper presents also the following direct-action passive safety components: modular pressure-relief safety valves; pilot safety valves with passive action; check valves with remote position indicator and after-tightening; modular inserts for limiting emergency coolant flow; vortex rectifier; critical weld fasteners; gas-liquid valves; fast-removable seal assembly; seal spring loaders; grooves for increasing hydraulic resistance. Replacement of active safety system components for passive ones improves the general reliability NPP by 1.5 or 2 orders of magnitudes. (authors)

  4. Manufacturing and testing experience for FFTF major safety related components

    Peckinpaugh, C.L.

    1976-01-01

    Experience with FFTF Heat Transport System components during design, manufacturing, and prototype testing is dscussed. Specifically the special design features and the results of the testing performed to assure that the designs provide for safe operation are outlined. Particular emphasis is placed on the full size prototype testing programs and the valuable experience gained

  5. Application of quality assurance program to safety related aging equipment or components

    Papaiya, N.C.

    1990-01-01

    This paper addresses how quality assurance programs and their criteria are applied to safety related and aging equipment or components used in commercial nuclear plant applications. The QA Programs referred to are 10CFR50 Appendix B and EPRI NP-5652. The QA programs as applicable are applied to equipment/component aging qualification, preventive maintenance, surveillance testing and procurement engineering. The intent of this paper is not the technical issues, methods and research of aging. The paper addresses QA program's application to age-related equipment or components in safety related applications. Quality Assurance Program 10CFR50 Appendix B applies to all safety related aging components or equipment related to the qualification program and associated preventive maintenance and surveillance testing programs. Quality Assurance involvement with procurement engineering for age-related commercial grade items supports EPRI NP-5652 and assures that the dedicated OGI is equal to the item purchased as a basic component to 10CFR50 Appendix B requirements

  6. A cost summary applicable to seismic construction and maintenance of nuclear safety related piping

    Stevenson, J.D.

    1995-01-01

    This paper presents a summary of costs applicable to nuclear power plant piping for an earthquake defined as 0.2 SSE-PGA as a function of three eras of initial construction: 1967--1974, 1974--1981 and 1981--1990. Costs have been presented for both new construction and maintenance in operating plants using both the original PSAR-FSAR design criteria and current SRP requirements. It is recommended that the cost information contained in this report be considered in evaluating the cost benefit relationships associated with current and proposed future changes in seismic design procedures applicable to safety-related piping systems

  7. Regulatory instrument review: Management of aging of LWR [light water reactor] major safety-related components

    Werry, E.V.

    1990-10-01

    This report comprises Volume 1 of a review of US nuclear plant regulatory instruments to determine the amount and kind of information they contain on managing the aging of safety-related components in US nuclear power plants. The review was conducted for the US Nuclear Regulatory Commission (NRC) by the Pacific Northwest Laboratory (PNL) under the NRC Nuclear Plant Aging Research (NPAR) Program. Eight selected regulatory instruments, e.g., NRC Regulatory Guides and the Code of Federal Regulations, were reviewed for safety-related information on five selected components: reactor pressure vessels, steam generators, primary piping, pressurizers, and emergency diesel generators. Volume 2 will be concluded in FY 1991 and will also cover selected major safety-related components, e.g., pumps, valves and cables. The focus of the review was on 26 NPAR-defined safety-related aging issues, including examination, inspection, and maintenance and repair; excessive/harsh testing; and irradiation embrittlement. The major conclusion of the review is that safety-related regulatory instruments do provide implicit guidance for aging management, but include little explicit guidance. The major recommendation is that the instruments be revised or augmented to explicitly address the management of aging

  8. Failure modes of safety-related components at fires on nuclear power plants

    Aaslund, A.

    2000-03-01

    Probabilistic assessment methods can be used to identify specific plant vulnerabilities. Application of such methods can also facilitate selection among system design alternatives available for safety enhancements. The quality of assessment results is however strongly dependent on realistic and accurate input data for modelling of system component behaviour and failure modes during conditions to be assessed. Use of conservative input data may not lead to results providing guidance on safety upgrades. Adequate input data for probabilistic assessments seems to be lacking for at least failure modes of some electrical components when exposed to a fire. This report presents an attempt to improve the situation with respect to such input data. In order to take advantage of information in existing documentation of fire incident occurrences some of the lessons learned from the fire at Browns Ferry Nuclear Power Plant on March 22, 1975 are discussed in this report. Also a summary of results from different fire tests of electrical cables presented in a fire risk analysis report is a part of the references. The failure modes used to describe fire-induced damage are 'open circuit' and 'hot short' which seems to be commonly accepted terms within the branch. Definitions of the terms are included in the report. Effects of the failure modes when occurring in some of the channels of the reactor protection system are discussed with respect to the existing design of the reactor protection system at Ringhals 2 nuclear power unit. Experiences from the Browns Ferry fire and results from fire tests of electrical cables indicate that the dominating failure mode for electrical cables is 'open circuit'. An 'open circuit' failure leads to circuit disjunction and loss of continuity. The circuit can no longer transmit its signal or power. When affecting channels of the reactor protection system an 'open circuit' failure can cause extensive inadvertent actions of safety related equipment

  9. Advancements in the design of safety-related systems and components of the MARS nuclear plant

    Caira, M.; Caruso, G.; Naviglio, A.; Sorabella, L.; Farello, C.E.

    1992-01-01

    In the paper, the advancements in the design of safety-related systems and components of the MARS nuclear plant, equipped with a 600 MW th PWR, are described. These advancements are due to the special safety features of this plant, which relies completely on inherent and passive safety. In particular, the new steps of the design of the innovative, completely passive, and with an unlimited autonomy Emergency core Cooling System are described, together with the characteristics of the last version of the steam generator, developed in a new design involving disconnecting components, for a fast erection and an easy maintenance. (author)

  10. Quality Control Activities Related to Mechanical Maintenance of Safety Related Components at Krsko NPP

    Djakovic, D.

    2016-01-01

    For successful, safe and reliable operation of nuclear power plant, maintenance processes have to be systematically controlled and procedures for quality control of maintenance activities shall be established. This is requested by the quality assurance program, which shall provide control over activities affecting the quality of structures, systems, and components, considering their importance to safety. As a part of Quality and Nuclear Oversight Division (QNOD; SKV), the Quality Control Department (QC) provides quality control activities, which are deeply involved in maintenance processes at Krsko NPP, both on safety related and non-safety related (non-nuclear safety) components. QC activities on safety related components have to fulfil all requirements, which will enable the components to perform their intended safety functions. This paper describes quality control activities related to mechanical maintenance of safety related components at Krsko NPP and significant role of the Krsko plant QC Department in three particular maintenance cases connected with safety related components. In these three specific cases, the QC has confirmed its importance in compliance with quality assurance program and presented its significant added value in providing safe and reliable operation of the plant. The first maintenance activity was installation of nozzle check valves in the scope of a modification for improving regulation of spent fuel pit pumps. The QC Department performed receipt inspection of the valves. Using non-destructive examination methods and X-ray spectrometry, it was found out that the valve diffuser was made of improper material, which could cause progressive corrosion of the valve diffuser in borated water and consequently a loss of safety function of the valves followed by long-term consequences. The second one was the receipt inspection of containment ventilation fan coolers. The coolers were claimed and sent back to the supplier because the QC Department

  11. Failure trend analysis for safety related components of Korean standard NPPs

    Choi, Sun Yeong; Han, Sang Hoon

    2005-01-01

    The component reliability data of Korean NPP that reflects the plant specific characteristics is required necessarily for PSA of Korean nuclear power plants. We have performed a project to develop the component reliability database (KIND, Korea Integrated Nuclear Reliability Database) and S/W for database management and component reliability analysis. Based on the system, we have collected the component operation data and failure/repair data during from plant operation date to 2002 for YGN 3, 4 and UCN 3, 4 plants. Recently, we provided the component failure rate data for UCN 3, 4 standard PSA model from the KIND. We evaluated the components that have high-ranking failure rates with the component reliability data from plant operation date to 1998 and 2000 for YGN 3,4 and UCN 3, 4 respectively. We also identified their failure mode that occurred frequently. In this study, we analyze the component failure trend and perform site comparison based on the generic data by using the component reliability data which is extended to 2002 for UCN 3, 4 and YGN 3, 4 respectively. We focus on the major safety related rotating components such as pump, EDG etc

  12. Optimal replacement policy for safety-related multi-component multi-state systems

    Xu Ming; Chen Tao; Yang Xianhui

    2012-01-01

    This paper investigates replacement scheduling for non-repairable safety-related systems (SRS) with multiple components and states. The aim is to determine the cost-minimizing time for replacing SRS while meeting the required safety. Traditionally, such scheduling decisions are made without considering the interaction between the SRS and the production system under protection, the interaction being essential to formulate the expected cost to be minimized. In this paper, the SRS is represented by a non-homogeneous continuous time Markov model, and its state distribution is evaluated with the aid of the universal generating function. Moreover, a structure function of SRS with recursive property is developed to evaluate the state distribution efficiently. These methods form the basis to derive an explicit expression of the expected system cost per unit time, and to determine the optimal time to replace the SRS. The proposed methodology is demonstrated through an illustrative example.

  13. The safety related aspects of pressure components in nuclear power plants

    Lindackers, K.H.

    1979-01-01

    Over the last two years the safety philosophy for nuclear power plants in the Federal Republic of Germany has changed considerably, as everyone working in the field perceives. The original and appropriate philosophy of risk minimalisation through graduated safety barriers has been more and more replaced by the utopian goal of total prevention of any damage. The reasons for this development are discussed briefly especially regarding pressure components. The very numerous pressure components of a nuclear power station are not all of equal importance with respect to safety. Although considerable efforts have been made, it has not been possible, to date, to achieve an agreement between operators, manufacturers, licensing authorities, independent experts, and other specialists about the safety related classification of the manifold pressure bearing parts in nuclear power stations. The background of this extremely regrettable situation is explained. In the last part of the paper the author suggests a simple and clear safety philosophy for pressure components in nuclear power stations. This philosophy is orientated both on Safety Regulations of the Radiation Protection Decree ('Strahlenschutzverordnung') of the 13th October 1976 and on the Safety Criteria for Nuclear Power Stations from 21st October 1977. Only a simple, clear framework can make a contribution to the further improvement of the already exceptional safety of nuclear facilities and to the removal of obstacles in the licensing procedure which, taken as a whole, tie up skilled personnel to a senseless degree, involve considerable financial expenditure, and have no relevance for the safety of nuclear power plants. (orig.) [de

  14. Development of reliability database for safety-related I and C component based on operating experience of KSNP

    Jang, S. C.; Han, S. H.; Min, K. R.

    2001-01-01

    Reliability database for safety-related I and C components has been developed, based on domestic operating experience of total 8.63 years from four units-Yonggwang Units 3 and 4, and Ulchin Units 3 and 4. This plant-specific data of safety-related I and C components has compared with operating experience for CE-supplied plants in U.S.A. As a results, we found that on the whole the domestic reliability data was similar to CE-supplied plants in USA, through lots of failures occurred early in the commercial operation were included in our analyses without percolation

  15. Environmental qualification - walkdowns: The documentation of configuration information for safety related components, equipment and systems

    Melmer, J.; Waters, M.

    1995-01-01

    Environmental Qualification walkdowns are conducted to collect field data to verify/validate/document configurations of safety related equipment and systems. This paper describes the process for conducting walkdowns and the justification for using an electronic format. The following are described: a) Background; b) Preparing, executing and processing walkdowns; c) Hardware/software; d) Impact of a paperless system on walkdown execution, maintenance and work planning; e) Other applications for the technology

  16. Common cause failure data collection and analysis for safety-related components of TRIGA SSR-14MW Pitesti, Romania

    Radu, G.; Mladin, D.

    2003-01-01

    This paper presents a study performed on the set of common cause failures (CCF) of safety-related components of the research reactor TRIGA SSR-14 MW Pitesti. The data collected cover a period of 20 years, from 1979 to 2000. The sources of data are Shift Supervisor Reports, Work Authorizations, and Reactor Log Books. Events collected are analyzed by failure mode and degrees of failure. Qualitative analysis of root causes, coupling factors and corrective actions and quantitative analysis of CCF events are studied. The objective of this work is to develop qualitative insights in the nature of the reported events and to build a site-specific common cause events database. (author)

  17. Literature review of environmental qualification of safety-related electric cables: Summary of past work. Volume 1

    Subudhi, M.

    1996-04-01

    This report summarizes the findings from a review of published documents dealing with research on the environmental qualification of safety-related electric cables used in nuclear power plants. Simulations of accelerated aging and accident conditions are important considerations in qualifying the cables. Significant research in these two areas has been performed in the US and abroad. The results from studies in France, Germany, and Japan are described in this report. In recent years, the development of methods to monitor the condition of cables has received special attention. Tests involving chemical and physical examination of cable's insulation and jacket materials, and electrical measurements of the insulation properties of cables are discussed. Although there have been significant advances in many areas, there is no single method which can provide the necessary information about the condition of a cable currently in service. However, it is possible that further research may identify a combination of several methods that can adequately characterize the cable's condition

  18. Literature review of environmental qualification of safety-related electric cables: Summary of past work. Volume 1

    Subudhi, M. [Brookhaven National Lab., Upton, NY (United States)

    1996-04-01

    This report summarizes the findings from a review of published documents dealing with research on the environmental qualification of safety-related electric cables used in nuclear power plants. Simulations of accelerated aging and accident conditions are important considerations in qualifying the cables. Significant research in these two areas has been performed in the US and abroad. The results from studies in France, Germany, and Japan are described in this report. In recent years, the development of methods to monitor the condition of cables has received special attention. Tests involving chemical and physical examination of cable`s insulation and jacket materials, and electrical measurements of the insulation properties of cables are discussed. Although there have been significant advances in many areas, there is no single method which can provide the necessary information about the condition of a cable currently in service. However, it is possible that further research may identify a combination of several methods that can adequately characterize the cable`s condition.

  19. Correlation of horizontal and vertical components of strong ground motion for response-history analysis of safety-related nuclear facilities

    Huang, Yin-Nan, E-mail: ynhuang@ntu.edu.tw [Dept. of Civil Engineering, National Taiwan University, No. 1, Sec. 4, Roosevelt Rd., Taipei 10617, Taiwan (China); Yen, Wen-Yi, E-mail: b01501059@ntu.edu.tw [Dept. of Civil Engineering, National Taiwan University, No. 1, Sec. 4, Roosevelt Rd., Taipei 10617, Taiwan (China); Whittaker, Andrew S., E-mail: awhittak@buffalo.edu [Dept. of Civil, Structural and Environmental Engineering, MCEER, State University of New York at Buffalo, Buffalo, NY 14260 (United States)

    2016-12-15

    Highlights: • The correlation of components of ground motion is studied using 1689 sets of records. • The data support an upper bound of 0.3 on the correlation coefficient. • The data support the related requirement in the upcoming edition of ASCE Standard 4. - Abstract: Design standards for safety-related nuclear facilities such as ASCE Standard 4-98 and ASCE Standard 43-05 require the correlation coefficient for two orthogonal components of ground motions for response-history analysis to be less than 0.3. The technical basis of this requirement was developed by Hadjian three decades ago using 50 pairs of recorded ground motions that were available at that time. In this study, correlation coefficients for (1) two horizontal components, and (2) the vertical component and one horizontal component, of a set of ground motions are computed using records from a ground-motion database compiled recently for large-magnitude shallow crustal earthquakes. The impact of the orientation of the orthogonal horizontal components on the correlation coefficient of ground motions is discussed. The rules in the forthcoming edition of ASCE Standard 4 for the correlation of components in a set of ground motions are shown to be reasonable.

  20. Columbia River Component Data Evaluation Summary Report

    C.S. Cearlock

    2006-08-02

    The purpose of the Columbia River Component Data Compilation and Evaluation task was to compile, review, and evaluate existing information for constituents that may have been released to the Columbia River due to Hanford Site operations. Through this effort an extensive compilation of information pertaining to Hanford Site-related contaminants released to the Columbia River has been completed for almost 965 km of the river.

  1. Interoperability Assets for Patient Summary Components: A Gap Analysis.

    Heitmann, Kai U; Cangioli, Giorgio; Melgara, Marcello; Chronaki, Catherine

    2018-01-01

    The International Patient Summary (IPS) standards aim to define the specifications for a minimal and non-exhaustive Patient Summary, which is specialty-agnostic and condition-independent, but still clinically relevant. Meanwhile, health systems are developing and implementing their own variation of a patient summary while, the eHealth Digital Services Infrastructure (eHDSI) initiative is deploying patient summary services across countries in the Europe. In the spirit of co-creation, flexible governance, and continuous alignment advocated by eStandards, the Trillum-II initiative promotes adoption of the patient summary by engaging standards organizations, and interoperability practitioners in a community of practice for digital health to share best practices, tools, data, specifications, and experiences. This paper compares operational aspects of patient summaries in 14 case studies in Europe, the United States, and across the world, focusing on how patient summary components are used in practice, to promote alignment and joint understanding that will improve quality of standards and lower costs of interoperability.

  2. Safety-related control air systems

    Anon.

    1977-01-01

    This Standard applies to those portions of the control air system that furnish air required to support, control, or operate systems or portions of systems that are safety related in nuclear power plants. This Standard relates only to the air supply system(s) for safety-related air operated devices and does not apply to the safety-related air operated device or to air operated actuators for such devices. The objectives of this Standard are to provide (1) minimum system design requirements for equipment, piping, instruments, controls, and wiring that constitute the air supply system; and (2) the system and component testing and maintenance requirements

  3. Qualification of safety-related valve actuators

    Anon.

    1981-01-01

    This Standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. It may also be used to separately qualify actuator components. This Standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of all safety-related functions of power-driven valve actuators

  4. Long term effects of the environment on safety related electric components in a nuclear power station. State-of-the-art

    Spaang, K.

    1984-01-01

    This paper reports the first stage in a research project aimed at finding and evaluating suitable methods to qualify electric components used in nuclear power plants. This part of the research project is concerned with the technical standards now used. The information is obtained at international conferences, visits to the industry and institutions in the USA and from literature studies. (K.A.E.)

  5. Conference summaries

    1991-01-01

    This volume contains conference summaries for the 31. annual conference of the Canadian Nuclear Association and the 12. annual conference of the Canadian Nuclear Society. Topics of discussion include: reactor physics; thermalhydraulics; industrial irradiation; computer applications; fuel channel analysis; small reactors; severe accidents; fuel behaviour under accident conditions; reactor components, safety related computer software; nuclear fuel management; fuel behaviour and performance; reactor safety; reactor engineering; nuclear waste management; and, uranium mining and processing

  6. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

  7. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

  8. Aging and service wear of hydraulic and mechanical snubbers used on safety-related piping and components of nuclear power plants. Phase I study

    Bush, S H; Heasler, P G; Dodge, R E

    1986-02-01

    This report presents an overview of hydraulic and mechanical snubbers used on nuclear piping systems and components, based on information from the literature and other sources. The functions and functional requirements of snubbers are discussed. The real versus perceived need for snubbers is reviewed, based primarily on studies conducted by a Pressure Vessel Research Committee. Tests conducted to qualify snubbers, to accept them on a case-by-case basis, and to establish their fitness for continued operation are reviewed. This report had two primary purposes. The first was to assess the effects of various aging mechanisms on snubber operation. The second was to determine the efficacy of existing tests in determining the effects of aging and degradation mechanisms. These tests include breakaway force, drag force, velocity/ acceleration range for activation in tension or compression, release rates within specified tension/compression limits, and restricted thermal movement. The snubber operating experience was reviewed using licensee event reports and other historical data for a period of more than 10 years. Data were statistically analyzed using arbitrary snubber populations. Value-impact was considered in terms of exposure to a radioactive environment for examination/ testing and the influence of lost snubber function and subsequent testing program expansion on the costs and operation of a nuclear power plant. The implications of the observed trends were assessed; recommendations include modifying or improving examination and testing procedures to enhance snubber reliability. Optimization of snubber populations by selective removal of unnecessary snubbers was also considered. (author)

  9. Requirements to be taken into account when designing safety-related mechanical components conveying or containing pressurized fluid and classified as level 2 or 3

    1984-12-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The purpose of this RFS is to specify the requirements to be taken into account when designing mechanical components conveying or containing pressurized fluid and which are in safety class 2 or 3

  10. Surry Power Station, Units 1 and 2. Annual operating report: January--December 1977, volume I--introduction, summary of operating experience; changes, tests, experiments, and safety-related maintenance; effluent releases; data tabulations

    1978-01-01

    A chronological operating sequence including shutdowns and occurrences during the year which required load reductions or resulted in non-load related incidents is given. Data are presented concerning plant and procedure changes, tests, experiments, safety related maintenance, effluent releases and personnel radiation exposures

  11. Configuration control during maintenance of safety related equipment

    Irish, C.S.

    2001-01-01

    Possibly the most important aspect of performing maintenance of safety related equipment is maintaining the component's original design basis. Assuring that the repaired item will perform the same safety function within the original performance and equipment qualification parameters is commonly referred to as configuration control. Maintaining configuration control of a technologically current well documented item is easy. Unfortunately, this does not describe most safety related items requiring maintenance within the global nuclear industry. Items such as motors, transformers, metal clad switchgear (low and medium voltage circuit breakers), refrigeration compressors, and electronic components (i.e. circuit boards, power supplies, regulators, etc.) which routinely require repair have been in service for twenty plus years. As a result, finding replacement parts and or material to repair the items to the original condition is becoming more and more difficult. An added difficulty is the lack of original technical documentation available on the item which is being repaired. The lack of technical documentation makes it difficult to identify replacement material and parts when the original part or material is not available. The lack of documentation also makes it difficult to test the repaired item to make sure that the original configuration has been maintained after the repair. The presentation will discuss the details of repairing various items including motors, metal clad switchgear, refrigeration compressors and power supplies and the controls which are necessary to maintain the configuration of the original item. The discussion will include the Quality Assurance and engineering necessary to identify and evaluate replacement material and parts necessary to perform repairs on safety related equipment when the original material or part is not available. Examples of repairs which required different parts or materials than the original to be used in the repair will be

  12. Configuration control during maintenance of safety related equipment

    Irish, C.S.

    2003-01-01

    Possibly the most important aspect of performing maintenance of safety related equipment is maintaining the component's original design basis. Assuring that the repaired item will perform the same safety function within the original performance and equipment qualification parameters is commonly referred to as configuration control. Maintaining configuration control of a technologically current well documented item is easy. Unfortunately, this does not describe most safety related items requiring maintenance within the global nuclear industry. Items such as motors, transformers, metal clad switchgear (low and medium voltage circuit breakers), refrigeration compressors, and electronic components (i.e. circuit boards, power supplies, regulators, etc.) which routinely require repair have been in service for twenty plus years. As a result, finding replacement parts and or material to repair the items to the original condition is becoming more and more difficult. An added difficulty is the lack of original technical documentation available on the item which is being repaired. The lack of technical documentation makes it difficult to identify replacement material and parts when the original part or material is not available. The lack of documentation also makes it difficult to test the repaired item to make sure that the original configuration has been maintained after the repair. The presentation will discuss the details of repairing various items including motors, metal clad switchgear, refrigeration compressors and power supplies and the controls which are necessary to maintain the configuration of the original item. The discussion will include the Quality Assurance and engineering necessary to identify and evaluate replacement material and parts necessary to perform repairs on safety related equipment when the original material or part is not available. Examples of repairs which required different parts or materials than the original to be used in the repair will be

  13. Materials and Components Technology Division research summary, 1992

    1992-11-01

    The Materials and Components Technology Division (MCT) provides a research and development capability for the design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs related to nuclear energy support the development of the Integral Fast Reactor (IFR): life extension and accident analyses for light water reactors (LWRs); fuels development for research and test reactors; fusion reactor first-wall and blanket technology; and safe shipment of hazardous materials. MCT Conservation and Renewables programs include major efforts in high-temperature superconductivity, tribology, nondestructive evaluation (NDE), and thermal sciences. Fossil Energy Programs in MCT include materials development, NDE technology, and Instrumentation design. The division also has a complementary instrumentation effort in support of Arms Control Technology. Individual abstracts have been prepared for the database

  14. Materials and Components Technology Division research summary, 1991

    1991-04-01

    This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base

  15. Materials and Components Technology Division research summary, 1991

    1991-04-01

    This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base.

  16. Safety-related control air systems - approved 1977

    Anon.

    1978-01-01

    This standard applies to those portions of the control air system that furnish air required to support, control, or operate systems or portions of systems that are safety related in nuclear power plants. This standard relates only to the air supply system(s) for safety-related air operated devices and does not apply to the safety-related air operated device or to air operated actuators for such devices. The objectives of this standard are to provide (1) minimum system design requirements for equipment, piping, instruments, controls, and wiring that constitute the air supply system; and (2) the system and component testing and maintenance requirements

  17. Safety related terms for advanced nuclear plants

    1995-12-01

    The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

  18. Safety related terms for advanced nuclear plants

    1991-09-01

    The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

  19. Summary of component reliability data for probabilistic safety analysis of Korean standard nuclear power plant

    Choi, S. Y.; Han, S. H.

    2004-01-01

    The reliability data of Korean NPP that reflects the plant specific characteristics is necessary for PSA of Korean nuclear power plants. We have performed a study to develop the component reliability DB and S/W for component reliability analysis. Based on the system, we had have collected the component operation data and failure/repair data during plant operation data to 1998/2000 for YGN 3,4/UCN 3,4 respectively. Recently, we have upgraded the database by collecting additional data by 2002 for Korean standard nuclear power plants and performed component reliability analysis and Bayesian analysis again. In this paper, we supply the summary of component reliability data for probabilistic safety analysis of Korean standard nuclear power plant and describe the plant specific characteristics compared to the generic data

  20. The development of summary components for the Disablement in the Physically Active scale in collegiate athletes.

    Houston, Megan N; Hoch, Johanna M; Van Lunen, Bonnie L; Hoch, Matthew C

    2015-11-01

    The Disablement in the Physically Active scale (DPA) is a generic patient-reported outcome designed to evaluate constructs of disability in physically active populations. The purpose of this study was to analyze the DPA scale structure for summary components. Four hundred and fifty-six collegiate athletes completed a demographic form and the DPA. A principal component analysis (PCA) was conducted with oblique rotation. Factors with eigenvalues >1 that explained >5 % of the variance were retained. The PCA revealed a two-factor structure consistent with paradigms used to develop the original DPA. Items 1-12 loaded on Factors 1 and Items 13-16 loaded on Factor 2. Items 1-12 pertain to impairment, activity limitations, and participation restrictions. Items 13-16 address psychosocial and emotional well-being. Consideration of item content suggested Factor 1 concerned physical function, while Factor 2 concerned mental well-being. Thus, items clustered around Factor 1 and 2 were identified as physical (DPA-PSC) and mental (DPA-MSC) summary components, respectively. Together, the factors accounted for 65.1 % of the variance. The PCA revealed a two-factor structure for the DPA that resulted in DPA-PSC and DPA-MSC. Analyzing the DPA as separate constructs may provide distinct information that could help to prescribe treatment and rehabilitation strategies.

  1. Advances in safety related maintenance

    2000-03-01

    The maintenance of systems, structures and components in nuclear power plants (NPPs) plays an important role in assuring their safe and reliable operation. Worldwide, NPP maintenance managers are seeking to reduce overall maintenance costs while maintaining or improving the levels of safety and reliability. Thus, the issue of NPP maintenance is one of the most challenging aspects of nuclear power generation. There is a direct relation between safety and maintenance. While maintenance alone (apart from modifications) will not make a plant safer than its original design, deficient maintenance may result in either an increased number of transients and challenges to safety systems or reduced reliability and availability of safety systems. The confidence that NPP structures, systems and components will function as designed is ultimately based on programmes which monitor both their reliability and availability to perform their intended safety function. Because of this, approaches to monitor the effectiveness of maintenance are also necessary. An effective maintenance programme ensures that there is a balance between the improvement in component reliability to be achieved and the loss of component function due to maintenance downtime. This implies that the safety level of an NPP should not be adversely affected by maintenance performed during operation. The nuclear industry widely acknowledges the importance of maintenance in NPP safety and operation and therefore devotes great efforts to develop techniques, methods and tools to aid in maintenance planning, follow-up and optimization, and in assuring the effectiveness of maintenance

  2. Criteria for safety-related operator actions

    Gray, L.H.; Haas, P.M.

    1983-01-01

    The Safety-Related Operator Actions (SROA) Program was designed to provide information and data for use by NRC in assessing the performance of nuclear power plant (NPP) control room operators in responding to abnormal/emergency events. The primary effort involved collection and assessment of data from simulator training exercises and from historical records of abnormal/emergency events that have occurred in operating plants (field data). These data can be used to develop criteria for acceptability of the use of manual operator action for safety-related functions. Development of criteria for safety-related operator actions are considered

  3. Structural Aging Program to evaluate continued performance of safety-related concrete structures in nuclear power plants

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1994-01-01

    This report discusses the Structural Aging (SAG) Program which is being conducted at the Oak Ridge National Laboratory (ORNL) for the United States Nuclear Regulatory commission (USNRC). The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved technical bases for their continued service. The program is organized into three technical tasks: Materials Property Data Base, Structural Component Assessment/Repair Technologies, and Quantitative Methodology for continued Service Determinations. Objectives and a summary of recent accomplishments under each of these tasks are presented

  4. Structural Aging Program approach to providing an improved basis for aging management of safety-related concrete structures

    Naus, D.J.; Oland, C.B.; Ellingwood, B.

    1993-01-01

    The Structural Aging (SAG) Program is being conducted at the Oak Ridge National Laboratory (ORNL) for the United States Nuclear Regulatory Commission (USNRC). The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved technical bases for their continued service. The program is organized into four tasks: Program Management, Materials Property Data Base, Structural Component Assessment/Repair Technologies, and Quantitative Methodology for Continued Service Determinations. Objectives and a summary of recent accomplishments under each of these tasks are presented

  5. Critical Propulsion Components. Volume 1; Summary, Introduction, and Propulsion Systems Studies

    2005-01-01

    Several studies have concluded that a supersonic aircraft, if environmentally acceptable and economically viable, could successfully compete in the 21st century marketplace. However, before industry can commit to what is estimated as a 15 to 20 billion dollar investment, several barrier issues must be resolved. In an effort to address these barrier issues, NASA and Industry teamed to form the High-Speed Research (HSR) program. As part of this program, the Critical Propulsion Components (CPC) element was created and assigned the task of developing those propulsion component technologies necessary to: (1) reduce cruise emissions by a factor of 10 and (2) meet the ever-increasing airport noise restrictions with an economically viable propulsion system. The CPC-identified critical components were ultra-low emission combustors, low-noise/high-performance exhaust nozzles, low-noise fans, and stable/high-performance inlets. Propulsion cycle studies (coordinated with NASA Langley Research Center sponsored airplane studies) were conducted throughout this CPC program to help evaluate candidate components and select the best concepts for the more complex and larger scale research efforts. The propulsion cycle and components ultimately selected were a mixed-flow turbofan (MFTF) engine employing a lean, premixed, prevaporized (LPP) combustor coupled to a two-dimensional mixed compression inlet and a two-dimensional mixer/ejector nozzle. Due to the large amount of material presented in this report, it was prepared in four volumes; Volume 1: Summary, Introduction, and Propulsion System Studies, Volume 2: Combustor, Volume 3: Exhaust Nozzle, and Volume 4: Inlet and Fan/ Inlet Acoustic Team.

  6. Summary

    2004-01-01

    The fourth workshop of the OECD/NEA Forum on Stakeholder Confidence (FSC) was hosted by ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste Management and enriched fissile materials. The central theme of the workshop was ''Dealing with interests, values and knowledge in managing risk''within the Belgian context of local partnerships for the long term management of low-level, short-lived radioactive waste. The four-day workshop started with a half-day session in Brussels giving a general introduction on the Belgian context and the local partnership methodology. This was followed by community visits to three local partnerships, PaLoFF in Fleurus-Farciennes, MONA in Mol, and STOLA in Dessel. After the visits, the workshop continued with two full-day sessions in Brussels. One hundred and nineteen registered participants, representing 13 countries, attended the workshop or participated in the community visits. About two thirds were Belgian stakeholders; the remainder came from FSC member organisations. The participants included representatives of municipal governments, civil society organisations, government agencies, industrial companies, the media, and international organisations as well as private citizens, consultants and academics. This Executive Summary gives an overview of the presentations and discussions that took place at the workshop and the community visits. The structure of the Executive Summary follows the structure of the workshop itself. Complementary to this Executive Summary and also provided with this document, is a NEA Secretariat's reflection aiming to place the main lessons of the workshop into an international perspective. (author)

  7. Questionnaire: involved actors in large disused components management - Summary Of Responses To The Questionnaire

    2012-01-01

    The aim of the Questionnaire is to establish an overview of the various bodies [Actors] that have responsibilities or input to the issue of large component decommissioning. In answering the intent is to cover the overall organisation and those bits that have most relevance to large components. The answers should reflect the areas from site operations to decommissioning as well as the wider issue of disposal at another location. The Questionnaire covers the following points: 1 - What is the country (institutional) structure for decommissioning? 2 - who does what and where lie the responsibilities? 3 - Which bodies have responsibility for onsite safety regulation, discharges and disposal? 4 - Which body(s) owns the facilities? 5 - Describe the responsibilities for funding of the decommissioning plan and disposal plan. Are they one and the same body? Whilst there are differences between countries there are some common threads. Regulation is through the state though the number of regulators involved may vary. In summary, the IAEA principles concerning independence of the regulatory body are followed. Funding arrangements vary but there are plans. Similarly, ownership of facilities is a mix of state and private. Some systems require a separate decommissioning license with Spain having the clearest demarcation of responsibilities for the decommissioning phase and waste management responsibilities

  8. Assessment and management of aging of nuclear power plant safety-related structures

    Naus, D.J.; Graves, H.L. III; Ellingwood, B.R.

    2003-01-01

    Background information and data have been developed for improving existing and developing new methods to assist in quantifying the effects of age-related degradation on the performance of nuclear power plant (NPP) safety-related structures. Factors that can lead to age-related degradation of safety-related structures are identified and their manifestations described. Current regulatory testing and inspection requirements are reviewed and a summary of degradation experience presented. Techniques commonly used to inspect NPP concrete structures to assess and quantify age-related degradation are summarized. An approach for conduct of condition assessments of structures in NPPs is presented. Criteria, based primarily on visual indications, are provided for use in classification and assessment of concrete degradation. Materials and techniques for repair of degraded structures are noted and guidance provided on repair options available for various forms of degradation. A probabilistic methodology for condition assessment and reliability-based life prediction has been developed and applied to structures subject to combinations of structural load processes and to structural systems. The methodology has also been used to investigate optimization of in-service inspection and maintenance strategies to maintain failure probability below a specified target value as well as to minimize costs. Fragility assessments involving analytical solutions and finite-element methods have been utilized to predict the effect of aging degradation on structural component performance. (author)

  9. Summary

    Roehlig, Klaus-Juergen

    2014-01-01

    and licensing process. This set of issues is by no means complete. For the Regulators' Forum and the IGSC it is now necessary to identify those issues and approaches to their resolutions which are of joint interest in order to address them in their programmes of work. The IGSC will, in accordance with its mandate, focus on topics related to safety case development and to the links to establish between different components of repository development. Subjects which have to be discussed and perhaps addressed in the Programme of Work include: - Operational safety: In the past, IGSC focussed on the relationship of operational and postclosure safety. A move towards questions specific for operational safety and in particular the potential for developing a list of events, incident causes etc. to be accounted for when assessing operational safety ('operational safety FEP list') will be considered. - Further attention will be devoted to establishing the linkage between the construction of engineered components and safety assessment, i.e. to the issue of feasibility to construct components according to the design specifications made by, or used in, safety assessments. - The IGSC will contribute to the EU MoDerN project in order to address issues related to monitoring and its linkage to safety demonstration. - IGSC also will further address organisational issues. (author)

  10. Summary

    1982-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-1 (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal

  11. Safety-related decision making at a nuclear power plant

    Vaurio, J.K.

    1998-01-01

    The decision making environment of an operating nuclear power plant is presented. The organizations involved, their roles and interactions as well as the main influencing factors and decision criteria are described. The focus is on safety-related decisions, and the framework is based on the situation at Loviisa power station. The role of probabilistic safety assessment (PSA) is illustrated with decisions concerning plant modifications, optimization, acceptance of temporary configurations and extended repair times. Suggestions are made for rational and flexible risk-based control of allowed times to operate the plant with some components out of service. (orig.)

  12. XUV synchrotron optical components for the Advanced Light Source: Summary of the requirements and the developmental program

    McKinney, W.; Irick, S.; Lunt, D.

    1992-07-01

    We give a brief summary of the requirements for water cooled optical components for the Advanced Light Source (ALS), a third generation synchrotron radiation source under construction at Lawrence Berkeley Laboratory (LBL). Materials choices, surface figure and smoothness specifications, and metrology systems for measuring the plated metal surfaces are discussed. Results from a finished water cooled copper alloy mirror will be used to demonstrate the state of the art in optical metrology with the Takacs Long Trace Profiler (LTP II)

  13. IEEE Std 382-1980: IEEE standard for qualification of safety-related valve actuators

    Anon.

    1992-01-01

    This standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. This standard may also be used to separately qualify actuator components. This standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of all safety-related functions of power-driven valve actuators

  14. Aging of nuclear safety related concrete structures

    Cerny, R.; Vydra, V.; Toman, J.; Vodak, F.

    1994-01-01

    An analysis of aging processes in nuclear-safety-related concrete structures (NSRCS) is presented. The major environmental stressor and aging factors affecting the performance of NSRCS are summarized, as are drying and plastic shrinkage, expansion of water during the freeze-thaw cycle, water passing through cracks dissolving or leaching the soluble calcium hydroxide, attack of acid rain and ground water, chemical reactions between particular aggregates and the alkaline solution within cement paste, reaction of calcium hydroxide in cement paste hydration products with atmospheric carbon dioxide, and physical radiation effects of neutrons and gamma radiation. The current methods for aging management in NSRCS are analyzed and evaluated. A new treatment is presented for the monitoring, evaluation and prediction of aging processes, consisting in a combination of theoretical methods, laboratory experiments, in-situ measurements and numerical simulations. 24 refs

  15. Development of safety related technology and infrastructure for safety assessment

    Venkat Raj, V.

    1997-01-01

    Development and optimum utilisation of any technology calls for the building up of the necessary infrastructure and backup facilities. This is particularly true for a developing country like India and more so for an advanced technology like nuclear technology. Right from the inception of its nuclear power programme, the Indian approach has been to develop adequate infrastructure in various areas such as design, construction, manufacture, installation, commissioning and safety assessment of nuclear plants. This paper deals with the development of safety related technology and the relevant infrastructure for safety assessment. A number of computer codes for safety assessment have been developed or adapted in the areas of thermal hydraulics, structural dynamics etc. These codes have undergone extensive validation through data generated in the experimental facilities set up in India as well as participation in international standard problem exercises. Side by side with the development of the tools for safety assessment, the development of safety related technology was also given equal importance. Many of the technologies required for the inspection, ageing assessment and estimation of the residual life of various components and equipment, particularly those having a bearing on safety, were developed. This paper highlights, briefly, the work carried out in some of the areas mentioned above. (author)

  16. 1st IAEA research coordination meeting on tritium retention in fusion reactor plasma facing components. October 5-6, 1995, Vienna, Austria. Summary report

    Langley, R.A.

    1995-12-01

    The proceedings and results of the 1st IAEA research Coordination Meeting on ''Tritium Retention in Fusion Reactor Plasma Facing Components'' held on October 5 and 6, 1995 at the IAEA Headquarters in Vienna are briefly described. This report includes a summary of presentations made by the meeting participants, the results of a data survey and needs assessment for the retention, release and removal of tritium from plasma facing components, a summary of data evaluation, and recommendations regarding future work. (author). 4 tabs

  17. Component Fragility Research Program: Phase 1, Demonstration tests: Volume 1, Summary report

    Holman, G.S.; Chou, C.K.; Shipway, G.D.; Glozman, V.

    1987-08-01

    This report describes tests performed in Phase I of the NRC Component Fragility Research Program. The purpose of these tests was to demonstrate procedures for characterizing the seismic fragility of a selected component, investigating how various parameters affect fragility, and finally using test data to develop practical fragility descriptions suitable for application in probabilistic risk assessments. A three-column motor control center housing motor controllers of various types and sizes as well as relays of different types and manufacturers was subjected to seismic input motions up to 2.5g zero period acceleration. To investigate the effect of base flexibility on the structural behavior of the MCC and on the functional behavior of the electrical devices, multiple tests were performed on each of four mounting configurations: four bolts per column with top bracking, four bolts per column with no top brace, four bolts per column with internal diagonal bracking, and two bolts per column with no top or internal bracking. Device fragility was characterized by contact chatter correlated to local in-cabinet response at the device location. Seismic capacities were developed for each device on the basis of local input motion required to cause chatter; these results were then applied to develop probabilistic fragility curves for each type of device, including estimates of the ''high-confidence low probability of failure'' capacity of each

  18. Materials Innovation for Next-Generation T&D Grid Components. Workshop Summary Report

    Taylor, Emmanuel [Energetics Incorporated, Columbia, MD (United States); Kramer, Caroline [Energetics Incorporated, Columbia, MD (United States); Marchionini, Brian [Energetics Incorporated, Columbia, MD (United States); Sabouni, Ridah [Energetics Incorporated, Columbia, MD (United States); Cheung, Kerry [U.S. Department of Energy (DOE), Washington, DC (United States). Office of Electricity Delivery and Energy Reliability (OE); Lee, Dominic F [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The Materials Innovations for Next-Generation T&D Grid Components Workshop was co-sponsored by the U.S. Department of Energy (DOE) Office of Electricity Delivery and Energy Reliability and the Oak Ridge National Laboratory (ORNL) and held on August 26 27, 2015, at the ORNL campus in Oak Ridge, Tennessee. The workshop was planned and executed under the direction of workshop co-chair Dr. Kerry Cheung (DOE) and co-chair Dr. Dominic Lee (ORNL). The information contained herein is based on the results of the workshop, which was attended by nearly 50 experts from government, industry, and academia. The research needs and pathways described in this report reflect the expert opinions of workshop participants, but they are not intended to represent the views of the entire electric power community.

  19. Safety-related parameters for the MAPLE research reactor and a comparison with the IAEA generic 10-MW research reactor

    Carlson, P.A.; Lee, A.G.; Smith, H.J.; Ellis, R.J.

    1989-07-01

    A summary is presented of some of the principle safety-related physics parameters for the MAPLE Research Reactor, and a comparison with the IAEA Generic 10-MW Reactor is given. This provides a means to assess the operating conditions and fuelling requirements for safe operation of the MAPLE Research Reactor under accepted standards

  20. 78 FR 29392 - Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied...

    2013-05-20

    ... NUCLEAR REGULATORY COMMISSION [NRC-2013-0098] Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied on for Safety AGENCY: Nuclear Regulatory Commission. ACTION... (NRC) is issuing for public comment Draft Regulatory Issue Summary (RIS) 2013-XX, ``Embedded Digital...

  1. RIBA Project - Risk-Informed approach for In-Service Inspection of Nuclear Power Plant Components. Project summary

    Lidbury, D.; Smith, G.

    2001-12-01

    The need for a European review of a Risk-Informed Approach for In-Service Inspection of Nuclear Power Plant Components (RIBA) was identified in 1998. This was as a priority item in the programme of activities conducted in the framework of the Council Resolutions of 22 July 1975 and of 18 June 1992 on the Technological Problems of Nuclear Safety. The RIBA Project was established in November 1999 as a 24-month Study Contract funded by the European Commission within the frame of the former DG XI WGCS (Working Group on Codes and Standards). The Study Contract was subsequently managed for the EC by DG TREN. The participants in RIBA were Serco Assurance (project coordinator), Ringhals AB, EDF, Tecnatom SA and Westinghouse Electric Europe. The work is presented in a summary report with the detailed results contained in three companion reports as follows: main conclusions and recommendations, Review of Existing Risk-Informed Methodologies, A Comparative Study of Risk-Informed In-Service Inspection Applications, Conclusions and Recommendations for Risk-Informed in-service inspection methodology applied to Nuclear Power Plants in Europe. (author)

  2. Analysis of factors affecting baseline SF-36 Mental Component Summary in Adult Spinal Deformity and its impact on surgical outcomes.

    Mmopelwa, Tiro; Ayhan, Selim; Yuksel, Selcen; Nabiyev, Vugar; Niyazi, Asli; Pellise, Ferran; Alanay, Ahmet; Sanchez Perez Grueso, Francisco Javier; Kleinstuck, Frank; Obeid, Ibrahim; Acaroglu, Emre

    2018-03-01

    To identify the factors that affect SF-36 mental component summary (MCS) in patients with adult spinal deformity (ASD) at the time of presentation, and to analyse the effect of SF-36 MCS on clinical outcomes in surgically treated patients. Prospectively collected data from a multicentric ASD database was analysed for baseline parameters. Then, the same database for surgically treated patients with a minimum of 1-year follow-up was analysed to see the effect of baseline SF-36 MCS on treatment results. A clinically useful SF-36 MCS was determined by ROC Curve analysis. A total of 229 patients with the baseline parameters were analysed. A strong correlation between SF-36 MCS and SRS-22, ODI, gender, and diagnosis were found (p baseline SF-36 MCS (p baseline SF-36 MCS in an ASD population are other HRQOL parameters such as SRS-22 and ODI as well as the baseline thoracic kyphosis and gender. This study has also demonstrated that baseline SF-36 MCS does not necessarily have any effect on the treatment results by surgery as assessed by SRS-22 or ODI. Level III, prognostic study. Copyright © 2018 Turkish Association of Orthopaedics and Traumatology. Production and hosting by Elsevier B.V. All rights reserved.

  3. Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

    Andermo, L.; Sundman, B.

    1974-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

  4. Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

    Olsson, S.; Andermo, L.

    1980-01-01

    This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

  5. Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

    Andermo, L.

    1977-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

  6. Feedback of safety - related operational experience: Lessons learned

    Elias, D [Commonwealth Edison Co. (United States)

    1997-09-01

    The presentation considers the following aspects of feedback of safety-related operational experience: lessons learned program, objectives, personnel characteristics; three types of documents for transmitting lessons learned issues.

  7. Feedback of safety - related operational experience: Lessons learned

    Elias, D.

    1997-01-01

    The presentation considers the following aspects of feedback of safety-related operational experience: lessons learned program, objectives, personnel characteristics; three types of documents for transmitting lessons learned issues

  8. Continuing the service of safety-related concrete structures in nuclear power plants

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.; Mori, Y.; Arndt, E.G.

    1993-01-01

    The Structural Aging (SAG) Program is addressing the aging management of safety-related concrete structures in nuclear power plants (NPPs) for the purpose of providing improved technical bases for their continued service. The program consists of three technical tasks: materials property data base, structural component assessment/repair technologies, and quantitative methodologies for continued service determinations. Recent accomplishments under each of these tasks are summarized

  9. Qualification of FPGA-Based Safety-Related PRM System

    Miyazaki, Tadashi; Oda, Naotaka; Goto, Yasushi; Hayashi, Toshifumi

    2011-01-01

    Toshiba has developed Non-rewritable (NRW) Field Programmable Gate Array (FPGA)-based safety-related Instrumentation and Control (I and C) system. Considering application to safety-related systems, nonvolatile and non-rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. FPGA is a device which consists only of basic logic circuits, and FPGA performs defined processing which is configured by connecting the basic logic circuit inside the FPGA. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing unit (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. The system which Toshiba developed this time is Power Range Neutron Monitor (PRM). Toshiba is planning to expand application of FPGA-based technology by adopting this development process to the other safety-related systems such as RPS from now on. Toshiba developed a special design process for NRW-FPGA-based safety-related I and C systems. The design process resolves issues for many years regarding testability of the digital system for nuclear safety application. Thus, Toshiba NRW-FPGA-based safety-related I and C systems has much advantage to be a would standard of the digital systems for nuclear safety application. (author)

  10. Safety related events at nuclear installations in 1995

    Korsbech, Uffe C C

    1996-01-01

    Nuclear safety related events of significance at least corresponding to level 2 of the International Nuclear Event Scale are described. In 1995 only two events occured at nuclear power plants, and four events occured at plants using ionizing radiation for processing or research.......Nuclear safety related events of significance at least corresponding to level 2 of the International Nuclear Event Scale are described. In 1995 only two events occured at nuclear power plants, and four events occured at plants using ionizing radiation for processing or research....

  11. Assessing propensity to learn from safety-related events

    Drupsteen, L.; Wybo, J.L.

    2015-01-01

    Most organisations aim to use experience from the past to improve safety, for instance through learning from safety-related incidents and accidents. Whether an organisation is able to learn successfully can however only be determined afterwards. So far, there are no proactive measures to assess

  12. Safety related analysis of the application and operation of electrical components in German nuclear power plants, safeguarding and protection against safety relevant impacts from the grid and other external sources; Sicherheitstechnische Analyse zum Einsatz und Betrieb elektrotechnischer Einrichtungen in deutschen Kernkraftwerken, Ueberwachung und Schutz gegen sicherheitstechnisch bedeutsame Einwirkungen aus dem Verbundnetz sowie anderen aeusseren Quellen

    Arians, Robert; Arnold, Simone; Blum, Stefanie; Buchholz, Marcel; Lochthofen, Andre; Quester, Claudia; Sommer, Dagmar

    2015-10-15

    In this report, results and data from examinations concerning software-based electrical components and transmitters are evaluated. As failure modes of software-based com-ponents and failure causes differ fundamentally from non-software-based components, an evaluation of the operating experience of such components was carried out. This evaluation should show whether or not existing approaches for non-software-based components can be directly transferred to software-based components, or if a different approach has to be developed. To include failures in non-safety systems, events not fulfilling the incident reporting criteria of German authorities were also included in this evaluation. The data provided by licensees of six German NPPs (different Boiling Wa-ter Reactors and Pressurized Water Reactors) was recorded for at least 8 years. The software-based components used in the NPPs are identified and their operating experience is analyzed in order to identify relevant failure modes and to establish a II knowledge base for future failure rating. In addition, the state of the art and science concerning the specific components was described.

  13. Summary record of the topical session at WPDD-10: Management of large components from decommissioning to storage and disposal, 18-19 November 2009

    Dutzer, Michel

    2010-01-01

    At its tenth meeting, the WPDD held a topical session on Management of Large Components from Decommissioning to Storage and Disposal. The topical session was organised by a new task group of the WPDD that recently began work on this topic. The group is aiming to prepare a technical guide that provides a methodology to assess different management options and facilitates involvement of the different interested parties in the process of selecting the preferred management option. This report is made of 3 parts: Part 1 presents the Main Outcomes of Topical Session on Management of Large Components from Decommissioning to Storage and Disposal (Summary of Presentations and Discussions and Rapporteurs Report); Part 2 presents the Agenda of the Topical Session on Management of Large Components from Decommissioning to Storage and Disposal; and Part 3 is the List of Participants

  14. Technical evaluation of seismic qualification of safety-related equipment

    Cho, Yang Hui; Park, Heong Gee; Park, Yeong Seok [Univ. of Incheon, Incheon (Korea, Republic of)

    1994-04-15

    This study is purposed to evaluate the technical acceptability of the procedures and techniques of seismic qualifications which were performed for the YGN 3 and 4 safety-related equipment.This study is also targeted to suggest a systematized technical procedure guide for the effective performance and review of the seismic qualification, which reflects the most up-to-date licensing requirements and state-of the-art.

  15. Safety-related operator actions: methodology for developing criteria

    Kozinsky, E.J.; Gray, L.H.; Beare, A.N.; Barks, D.B.; Gomer, F.E.

    1984-03-01

    This report presents a methodology for developing criteria for design evaluation of safety-related actions by nuclear power plant reactor operators, and identifies a supporting data base. It is the eleventh and final NUREG/CR Report on the Safety-Related Operator Actions Program, conducted by Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The operator performance data were developed from training simulator experiments involving operator responses to simulated scenarios of plant disturbances; from field data on events with similar scenarios; and from task analytic data. A conceptual model to integrate the data was developed and a computer simulation of the model was run, using the SAINT modeling language. Proposed is a quantitative predictive model of operator performance, the Operator Personnel Performance Simulation (OPPS) Model, driven by task requirements, information presentation, and system dynamics. The model output, a probability distribution of predicted time to correctly complete safety-related operator actions, provides data for objective evaluation of quantitative design criteria

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

    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)

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

    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)

  18. IEEE Std 382-1985: IEEE standard for qualification of actuators for power operated valve assemblies with safety-related functions for nuclear power plants

    Anon.

    1992-01-01

    This standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. This standard may also be used to separately qualify actuator components. This standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of power-driven valve actuators with safety-related functions Part I describes the qualification process. Part II describes the standard qualification cases and their environmental parameters for the usual locations of safety-related equipment in a nuclear generating station. Part III describes the qualification tests outlined in 6.3.3

  19. Safety-related occurrences at the Finnish nuclear power plants

    Reponen, H.; Viitasaari, O.

    1985-04-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Report for this period, Operation of Finnish Nuclear Power Plants (STUK-B-YTO 7), which is supplemented by this report intended for experts. (author)

  20. Modification and backfitting in safety related systems at Ringhals 2

    Lidh, B. [KSU, Nykoeping (Sweden); Stroemqvist, E. [ES-Konsult AB, Stockholm (Sweden)

    1995-08-01

    This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Ringhals-2, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 4 refs.

  1. Safety-related occurrences at the Finnish nuclear power plants

    Viitasaari, O.; Rantavaara, A.

    1984-03-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Report for this period, Operation of Finnish Nuclear Power Plants (STL-B-RTO-83/7), which is supplemented by this report intended principally for experts. (author)

  2. Safety-related incidents at the Finnish nuclear power plants

    Lehtinen, P.

    1985-01-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Reports, Operation of Finnish Nuclear Power Plants, which are supplemented by this report intended for experts. (author)

  3. Safety-related incidents at the Finnish nuclear power plants

    Lehtinen, P.

    1986-03-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Reports, Operation of Finnish Nuclear Power Plants, which are supplemented by this report intended for experts. (author)

  4. Modification and backfitting in safety related systems at Ringhals 2

    Lidh, B.; Stroemqvist, E.

    1995-08-01

    This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Ringhals-2, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 4 refs

  5. Reliability of containment and safety-related structures

    Nessim, M.A.

    1995-09-01

    A research program on Reliability of Containment and Safety-related Structures has been developed and is described in this document. This program is designed to support AECB's regulatory activities aimed at ensuring the safety of these structures. These activities include evaluating submissions by operators and requesting special assessments when necessary. The results of the proposed research will also be useful in revising and enhancing the CSA design standards for containment and safety-related structures. The process of developing the research program started with an information collection and review phase. The sources of information included C-FER's previous work in the area, various recent research publications, regulatory documents and relevant design standards, and a detailed discussion with AECB staff. The second step was to outline the process of reliability evaluation, and identify the required models and parameters. Comparison between the required and available information was used to identify gaps in the state-of-the-art, and the research program was designed to fill these gaps. The program is organized in four major topics, namely: development of an approach for reliability analysis; compilation and development of the required analysis tools; application to specific problems related to design, assessment, maintenance and testing of structures; and testing and validation. It is suggested that the program should be supported by an on-going process of communication and consultation between AECB staff and industry experts. This will lend credibility to the results and facilitate their future application. (author). 1 fig

  6. Design and installation of advanced computer safety related instrumentation

    Koch, S.; Andolina, K.; Ruether, J.

    1993-01-01

    The rapidly developing area of computer systems creates new opportunities for commercial utilities operating nuclear reactors to improve plant operation and efficiency. Two of the main obstacles to utilizing the new technology in safety-related applications is the current policy of the licensing agencies and the fear of decision making managers to introduce new technologies. Once these obstacles are overcome, advanced diagnostic systems, CRT-based displays, and advanced communication channels can improve plant operation considerably. The article discusses outstanding issues in the area of designing, qualifying, and licensing of computer-based instrumentation and control systems. The authors describe the experience gained in designing three safety-related systems, that include a Programmable Logic Controller (PLC) based Safeguard Load Sequencer for NSP Prairie Island, a digital Containment Isolation monitoring system for TVA Browns Ferry, and a study that was conducted for EPRI/NSP regarding a PLC-based Reactor Protection system. This article presents the benefits to be gained in replacing existing, outdated equipment with new advanced instrumentation

  7. Compiler issues associated with safety-related software

    Feinauer, L.R.

    1991-01-01

    A critical issue in the quality assurance of safety-related software is the ability of the software to produce identical results, independent of the host machine, operating system, or compiler version under which the software is installed. A study is performed using the VIPRE-0l, FREY-01, and RETRAN-02 safety-related codes. Results from an IBM 3083 computer are compared with results from a CYBER 860 computer. All three of the computer programs examined are written in FORTRAN; the VIPRE code uses the FORTRAN 66 compiler, whereas the FREY and RETRAN codes use the FORTRAN 77 compiler. Various compiler options are studied to determine their effect on the output between machines. Since the Control Data Corporation and IBM machines inherently represent numerical data differently, methods of producing equivalent accuracy of data representation were an important focus of the study. This paper identifies particular problems in the automatic double-precision option (AUTODBL) of the IBM FORTRAN 1.4.x series of compilers. The IBM FORTRAN version 2 compilers provide much more stable, reliable compilation for engineering software. Careful selection of compilers and compiler options can help guarantee identical results between different machines. To ensure reproducibility of results, the same compiler and compiler options should be used to install the program as were used in the development and testing of the program

  8. Safety related terms for advanced nuclear plants; Terminos relacionados con la seguridad para centrales nucleares avanzadas

    NONE

    1995-12-01

    The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety.

  9. Seismic design and performance of nuclear safety related RC structures based on new seismic design principle

    Murugan, R.; Sivathanu Pillai, C.; Chattopadhyaya, S.; Sundaramurthy, C.

    2011-01-01

    Full text: Seismic design of safety related Reinforced Concrete (RC) structures of Nuclear power plants (NPP) in India as per the present AERB codal procedures tries to ensure predominantly elastic behaviour under OBE so that the features of Nuclear Power Plant (NPP) necessary for continued safe operation are designed to remain functional and prevent accident (collapse) of NPP under SSE for which certain Structures, Systems and Components (SSCs) those are necessary to ensure the capability to shut down the reactor safely, are designed to remain functional. While the seismic design principles of non safety related structures as per Indian code (IS 1893-2002) are ensuring elastic behaviour under DBE and inelastic behaviour under MCE by utilizing ductility and energy dissipation capacity of the structure effectively. The design principle of AERB code is ensuring elastic behaviour under OBE and is not enlightening much inference about the overall structural behaviour under SSE (only ensuring the capability of certain SSCs required for safe shutdown of reactor). Various buildings and structures of Indian Nuclear power plant are classified from the basis of associated safety functions in a descending order in according with their roles in preventions and mitigation of an accident or support functions for prevention. This paper covers a comprehensive seismic analysis and design methodology based on the AERB codal provisions followed for safety related RC structure taking Diesel Generator Building of PFBR as a case study and study and investigates its performance under OBE and SSE by carrying out Non-linear static Pushover analysis. Based on the analysis, observed variations, recommendations are given for getting the desired performance level so as to implement performance based design in the future NPP design

  10. Safety design guide for safety related systems for CANDU 9

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Wright, A.C.D. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new.

  11. Value-impact assessment of safety-related modifications

    Knowles, W.M.C.; Dinnie, K.S.; Gordon, C.W.

    1992-01-01

    Like other nuclear utilities, Ontario Hydro, as part of its risk management activities, continually assesses the safety of its nuclear operations. In addition, new regulatory requirements are being applied to the older nuclear power plants. Both of these result in proposed plant modifications designed to reduce the risk to the public. However, modifications to an operating plant can have serious economic effects, and the resources, both financial and personnel, required for the implementation of these modifications are limited. Thus, all potential benefits and effects of a proposed modification must be thoroughly investigated to judge whether the modification is beneficial. Ontario Hydro has begun to use comprehensive value-impact assessments, utilizing plant-specific probabilistic risk assessments (PRAs), as tools to provide an informed basis for judgments on the benefit of safety-related modifications. The results from value-impact assessments can also be used to prioritize the implementation of these modifications

  12. Safety design guide for safety related systems for CANDU 9

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new

  13. Safety-related LWR research. Annual report 1989

    1990-11-01

    The main topics in this annual report 1989 are phenomena of heavy fuel damage and single aspects of a core meltdown accident. The examined single aspects refer to aerosol behavior and filter engineering and to methods for assessment and minimization of the radiological consequences of reactor accidents. Different contributions to selected, safety-related problems of an advanced pressurized-water reactor complete the topic spectrum. The annual report 1989 describes the progress of the research work wich was carried out in the area of safety research by institutes and departments of the KfK, and on behalf of the KfK by external institutions. The individual contributions represent the status of work at the end of the year under review, 1989. (orig./HP) [de

  14. Usage of Commercial Grade Programmable Digital Systems in Safety Related Applications

    Mandic, D.

    2006-01-01

    This paper explains methods and conditions, which if completely and correctly fulfilled, enable an operating NPP (Nuclear Power Plant) licensed and operating in accordance with the US codes and US regulatory requirements to use a commercial grade programmable digital device (PLC - Programmable Digital Controller, digital controller, digital computer or process computer) in a safety related application in a NPP. In mid 80's, when an intensive construction cycle of the new NPPs in the U.S.A. was completed, many equipment manufacturers either disappeared from the market or they abandoned their product lines that were designed and manufactured under 10 CFR Part 50 Appendix B quality assurance program. The quality assurance as defined by 10 CFR Part 50 Appendix B comprises all those planned and systematic actions necessary to provide adequate confidence that a Structure, System or Component (SSC) will perform satisfactorily in service . The operating NPPs faced the problem related to the availability of qualified equipment, components and spare parts. The US NRC (Nuclear Regulatory Commission) recognized that problem timely (Oct. 1978 revision of 10CFR21) and required a commercial grade item to be dedicated before it could be used as a basic component. A special process named Dedication of CGI - Commercial Grade Items if conducted properly, provides reasonable assurance that a commercial grade item to be used as a basic component will perform its intended safety related function and, in this respect, is deemed equivalent to an item designed and manufactured under 10 CFR Part 50 Appendix B. After that, the Dedication of CGI has been widely used mostly for relatively simple mechanical, electrical, and IandC components and spare parts. In order to provide guidance to the dedication process, EPRI has issued two documents (EPRI NP-5652 and Supplemental Guidance for EPRI NP-5652). All nuclear power plants, which comply with the US nuclear regulatory requirements, hindered as

  15. 77 FR 6411 - Training, Qualification, and Oversight for Safety-Related Railroad Employees

    2012-02-07

    ... Oversight for Safety-Related Railroad Employees AGENCY: Federal Railroad Administration (FRA), Department of... establishing minimum training standards for each category and subcategory of safety-related railroad employee... or contractor that employs one or more safety-related railroad employee to develop and submit a...

  16. A summary of recent refinements to the WAKE dispersion model, a component of the HGSYSTEM/UF6 model suite

    Yambert, M.W.; Lombardi, D.A.; Goode, W.D. Jr.; Bloom, S.G.

    1998-08-01

    The original WAKE dispersion model a component of the HGSYSTEM/UF 6 model suite, is based on Shell Research Ltd.'s HGSYSTEM Version 3.0 and was developed by the US Department of Energy for use in estimating downwind dispersion of materials due to accidental releases from gaseous diffusion plant (GDP) process buildings. The model is applicable to scenarios involving both ground-level and elevated releases into building wake cavities of non-reactive plumes that are either neutrally or positively buoyant. Over the 2-year period since its creation, the WAKE model has been used to perform consequence analyses for Safety Analysis Reports (SARs) associated with gaseous diffusion plants in Portsmouth (PORTS), Paducah (PGDP), and Oak Ridge. These applications have identified the need for additional model capabilities (such as the treatment of complex terrain and time-variant releases) not present in the original utilities which, in turn, has resulted in numerous modifications to these codes as well as the development of additional, stand-alone postprocessing utilities. Consequently, application of the model has become increasingly complex as the number of executable, input, and output files associated with a single model run has steadily grown. In response to these problems, a streamlined version of the WAKE model has been developed which integrates all calculations that are currently performed by the existing WAKE, and the various post-processing utilities. This report summarizes the efforts involved in developing this revised version of the WAKE model

  17. Challenges in the management of gas voids in safety related systems

    Ezekoye, L.I.; Turkowski, W.M.; Ferraraccio, F.P.; Swartz, M.M.

    2009-01-01

    Gas intrusion into Safety Related Systems, such as the Emergency Core Cooling System (ECCS), Decay Heat Removal (DHR) and Containment Spray (CS) in nuclear power plants is undesirable and can lead to pump binding (depending on the void fraction and flow rate) and damaging water hammer events. Gas ingestion in pumps can result in total or momentary loss of hydraulic performance resulting in possible pump shaft seizure rendering the pumps unable to perform their safety functions or reduce the pump discharge pressure and flow capacity to the point that the system cannot perform its design function. Extreme cases of gas water hammer can result in physical damage to system piping, components and supports, and possible relief valve lifting events with consequential loss of inventory. NRC Generic Letter GL 2008 01, 'Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,' requires US utilities to demonstrate that suitable design, operational and testing measures are in place to maintain licensing commitments. The Generic Letter (GL 2008 01) outlines a number of actions that are detailed in nature, such as establishing pump void tolerance limits; establishing limits on pump suction void fractions, assuring adequate system venting capability, identification of all possible sources of gas intrusion, preventing vortex formation in tanks, and determining acceptable limits of gas in system discharge piping.. Regarding one of these issues, GL 2008 01 indicates that the amount of gas that can be ingested without significant impact on pump design, gas dispersion and flow rate. Each US nuclear power plant licensee is required to evaluate their ECCS, DHR and CS system design, operation and test procedures to assure that gas intrusion is minimized and monitored in order to maintain system operability and compliance with the requirements of 10 CFR 50 Appendix B. Typically, gas pockets get into the safety related systems through a number

  18. Challenges in the management of gas voids in safety related systems

    Ezekoye, L.I.; Turkowski, W.M.; Ferraraccio, F.P.; Swartz, M.M. [Westinghouse Electric Company LLC, Pittsburgh (United States)

    2009-04-15

    Gas intrusion into Safety Related Systems, such as the Emergency Core Cooling System (ECCS), Decay Heat Removal (DHR) and Containment Spray (CS) in nuclear power plants is undesirable and can lead to pump binding (depending on the void fraction and flow rate) and damaging water hammer events. Gas ingestion in pumps can result in total or momentary loss of hydraulic performance resulting in possible pump shaft seizure rendering the pumps unable to perform their safety functions or reduce the pump discharge pressure and flow capacity to the point that the system cannot perform its design function. Extreme cases of gas water hammer can result in physical damage to system piping, components and supports, and possible relief valve lifting events with consequential loss of inventory. NRC Generic Letter GL 2008 01, 'Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,' requires US utilities to demonstrate that suitable design, operational and testing measures are in place to maintain licensing commitments. The Generic Letter (GL 2008 01) outlines a number of actions that are detailed in nature, such as establishing pump void tolerance limits; establishing limits on pump suction void fractions, assuring adequate system venting capability, identification of all possible sources of gas intrusion, preventing vortex formation in tanks, and determining acceptable limits of gas in system discharge piping.. Regarding one of these issues, GL 2008 01 indicates that the amount of gas that can be ingested without significant impact on pump design, gas dispersion and flow rate. Each US nuclear power plant licensee is required to evaluate their ECCS, DHR and CS system design, operation and test procedures to assure that gas intrusion is minimized and monitored in order to maintain system operability and compliance with the requirements of 10 CFR 50 Appendix B. Typically, gas pockets get into the safety related systems through

  19. Aging of turbine drives for safety-related pumps in nuclear power plants

    Cox, D.F.

    1995-06-01

    This study was performed to examine the relationship between time-dependent degradation and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety-related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized-water reactor plants and in the Reactor Core Isolation Cooling and High-Pressure Coolant Injection systems for boiling-water reactor plants. This research has been conducted by examination of failure data in the Nuclear Plant Reliability Data System, review of Licensee Event Reports, discussion of problems with operating plant personnel, and personal observation. The reported failure data were reviewed to determine the cause of the event and the method of discovery. Based on the research results, attempts have been made to determine the predictability of failures and possible preventive measures that may be implemented. Findings in a recent study of AFW systems indicate that the turbine drive is the single largest contributor to AFW system degradation. However, examination of the data shows that the turbine itself is a reliable piece of equipment with a good service record. Most of the problems documented are the result of problems with the turbine controls and the mechanical overspeed trip mechanism; these apparently stem from three major causes which are discussed in the text. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design, have led to improved performance resulting in a reliable safety-related component. However, these improvements have not been universally implemented

  20. Software FMEA analysis for safety-related application software

    Park, Gee-Yong; Kim, Dong Hoon; Lee, Dong Young

    2014-01-01

    Highlights: • We develop a modified FMEA analysis suited for applying to software architecture. • A template for failure modes on a specific software language is established. • A detailed-level software FMEA analysis on nuclear safety software is presented. - Abstract: A method of a software safety analysis is described in this paper for safety-related application software. The target software system is a software code installed at an Automatic Test and Interface Processor (ATIP) in a digital reactor protection system (DRPS). For the ATIP software safety analysis, at first, an overall safety or hazard analysis is performed over the software architecture and modules, and then a detailed safety analysis based on the software FMEA (Failure Modes and Effect Analysis) method is applied to the ATIP program. For an efficient analysis, the software FMEA analysis is carried out based on the so-called failure-mode template extracted from the function blocks used in the function block diagram (FBD) for the ATIP software. The software safety analysis by the software FMEA analysis, being applied to the ATIP software code, which has been integrated and passed through a very rigorous system test procedure, is proven to be able to provide very valuable results (i.e., software defects) that could not be identified during various system tests

  1. Can we use IEC 61850 for safety related functions?

    Luca Rocca

    2016-08-01

    Full Text Available Safety is an essential issue for processes that present high risk for human beings and environment. An acceptable level of risk is obtained both with actions on the process itself (risk reduction and with the use of special safety systems that switch the process into safe mode when a fault or an abnormal operation mode happens. These safety systems are today based on digital devices that communicate through digital networks. The IEC 61508 series specifies the safety requirements of all the devices that are involved in a safety function, including the communication network. Also electrical generation and distribution systems are processes that may have a significant level of risk, so the criteria stated by the IEC 61508 applies. Starting from this consideration, the paper analyzes the safety requirement for the communication network and compare them with the services of the communication protocol IEC 61850 that represents the most used protocol for automation of electrical plants. The goal of this job is to demonstrate that, from the technical point of view, IEC 61850 can be used for implementing safety-related functions, even if a formal safety certification is still missing.

  2. Methods used to seismically upgrade. The safety related components of Belgian plants

    Lafaille, J.P.

    1993-01-01

    Belgian nuclear power amounts to about 6,000 MW, generated by seven plants that started operation as early as 1967. The latest plant started in 1985. Some of these plants were designed with no seismic requirements whatsoever. Even for those that had seismic requirements at the design stage, seismic demand was raised after design had been frozen (late during construction or at the 10 years revision). As a consequence all the plants had to undergo, to a variable extent, a seismic reevaluation and/or backfitting. Civil structures were concerned as well as electro-mechanical equipment and piping systems. The present paper deals with the mechanical aspect of the problem (equipment and piping). In order to minimize hardware modifications, advanced analytical techniques were used throughout the process, starting with the elaboration of a site specific spectrum, and using a full soil-structure interaction in order to get as 'realistic' as possible floor response spectra. In some instances, non linear elasto-plastic time history analysis was performed on piping-systems in order to qualify them without hardware modifications. In other cases a 'Load Coefficient Method' was used. Sometimes stresses or displacements taken from the original stress reports and scaled by comparison of applicable spectra, allowed to assess the seismic validity of the system under investigation. Seismic acceptability of installed active equipment is more difficult to demonstrate, as this is usually done by testing. This problem is a generic issue in the US, identified under the label USI-A-46 (Unresolved Safety Issue). It is treated by. a group of Utilities (SQUG = Seismic Qualification Utilities Group). The Belgian Utility is member of that group since 1985. The application of this program is starting in the US. SQUG methodology has been applied to three Belgian plants starting in 1988 and is now completed. The required fixes are being implemented. Experience gained in the process has been applied to other Belgian units. General conclusions can be drawn. They are presented in the paper. Among them is the necessity of performing 'seismic markdowns' at regular intervals during the plant life in order to restore the seismic adequacy of vital equipment possibly degraded by 'maintenance drift'. Among the required seismic upgrades is the necessity, in a near future, to define accurate measures to be taken in the case of occurrence of an earthquake close to the OBE level. It has become clear that the input for defining these actions is readily available as a result of the SQUG walkdowns. (author)

  3. Quality of Life in Persons Living With an Ostomy Assessed Using the SF36v2: Mental Component Summary: Vitality, Social Function, Role-Emotional, and Mental Health.

    Nichols, Thom R

    The purpose of this study was to assess the Mental Health Component of health-related quality of life (HRQOL) in community-dwelling persons with ostomies residing in the United States. Cross-sectional descriptive study. Two thousand three hundred twenty-nine participants completed the survey for a response rate of 14.9% and a margin of error of 2.03%. Study respondents were geographically distributed throughout the United States, representing all 50 states. Fifty-three percent of study respondents were male. Respondents had a median age of 65 years. Forty percent have colostomies, 44% are living with ileostomies, and 13% have urostomies. The remaining 3% are living with multiple stomas or they indicated that they were uncertain as to the type of stoma. The SF36v2 was used to assess HRQOL. This instrument was selected because it has the ability to measure HRQOL in a target population and it allows comparison with the general population. Potential participants were randomly selected from an electronic database of 15,591 persons with ostomies. They were contacted by e-mails and provided with an electronic nontransferable link to the survey. This is a secondary analysis of findings from the Mental Component Summary (MCS) of the SF36v2. Persons who have undergone ostomy surgery did not score as well as the general population when components of the MCS were compared. While overall differences were identified, they differed based on age and cumulative MCS score levels. Analysis of individuals found to have significant impairment in MCS scores (cumulative soccer ostomies as lower than scores generated from the general population. However, these findings varied based on age and cumulative MCS score.

  4. Categorization of safety related motor operated valve safety significance for Ulchin Unit 3

    Kang, D. I.; Kim, K. Y.

    2002-03-01

    We performed a categorization of safety related Motor Operated Valve (MOV) safety significance for Ulchin Unit 3. The safety evaluation of MOV of domestic nuclear power plants affects the generic data used for the quantification of MOV common cause failure ( CCF) events in Ulchin Units 3 PSA. Therefore, in this study, we re-estimated the MGL(Multiple Greek Letter) parameter used for the evaluation of MOV CCF probabilities in Ulchin Units 3 Probabilistic Safety Assessment (PSA) and performed a classification of the MOV safety significance. The re-estimation results of the MGL parameter show that its value is decreased by 30% compared with the current value in Ulchin Unit 3 PSA. The categorization results of MOV safety significance using the changed value of MGL parameter shows that the number of HSSCs(High Safety Significant Components) is decreased by 54.5% compared with those using the current value of it in Ulchin Units 3 PSA

  5. Towards assuring the continued performance of safety-related concrete structures in nuclear power plants

    Naus, D.J.; Oland, C.B.; Ellingwood, B.; Mori, Y.; Arndt, E.G.

    1993-01-01

    The Structural Aging (SAG) Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved technical bases for their continued service. Pertinent concrete structures are described in terms of their importance, design considerations, and materials of construction. Degradation factors which can potentially impact the ability of these structures to meet their functional and performance requirements are identified. A review of the performance history of the concrete components in nuclear power plants is provided. Accomplishments of the SLAG Program are summarized, i.e., development of the structural materials information center, development of a structural aging assessment methodology, evaluation of models for predicting the remaining life of in-service concrete, review of in-service inspection methods, and development of a methodology for reliability-based condition assessment and life prediction of concrete structures. On-going activities are also described

  6. Integrity of Safety-Related Fast Reactor Structures

    Rose, R.T.; Tomkins, B.

    1981-01-01

    The LMFBR contains several structural items whose integrity must be safeguarded during the life of the plant. These items include the main core support structures (strongback, diagrid) and the primary tank to which these structures are attached. In order to demonstrate an acceptable level of structural integrity, the chosen design philosophy must be supported by both analytical and experimental evidence. This paper describes the current approaches in the UK to these requirements. Section 2 describes the materials mechanical properties tests performed to date on both fracture toughness and fatigue crack growth in Type 316 austenitic stainless steel plate and weldments. This data illustrates the problems in identifying the relevant materials fracture parameters for use in assessments. Section 3 shows the test programmes in hand to extend the materials programmes to tests on structural features (mainly welded wide plate tests) which incorporate the complexity of weldments in a structural context. This includes experimental evidence on the effects of local weld residual stresses on structural failure. Various routes are open for the integrity assessment of FR structures. These are discussed in Section 4 but in effect they reduce to a fracture mechanics approach using some technique to cope with elastic-plastic fracture. The main problems at present relate to our ability in analysis to cope with residual stresses and the post-initiation region of the fracture resistance curve. Also, there is the problem of initial defect sizing by current NDE techniques. Current conservative analytical assessments give acceptable defect sizes of order a few millimetres in irradiated weldments. Finally, Section 5 discusses the options open in design to cope with safety related structures under normal and abnormal loading conditions. It is clear that several options exist in design to satisfy the demand for high integrity

  7. Benefits of a systematic approach to maintenance for safety and safety related systems

    Dam, R.F.; Ayazzudin, S.; Nickerson, J.H.

    2003-01-01

    For safety and safety-related systems, nuclear plants have to balance the requirements of demonstrating the reliability of each system, while maintaining the system and plant availability. With the goal of demonstrating statistical reliability, these systems have extensive testing programs, which often results in system unavailability and this can impact the plant capacity. The inputs to the process are often safety and regulatory related, resulting in programs that provide a high level of scrutiny. In such cases, the value of the application of a Systematic Assessment of Maintenance (SAM) process, such as Reliability Centered Maintenance (RCM), is questioned. The special case of Standby-Safety systems was discussed in a previous paper, where it was demonstrated how SAM techniques provide useful insight into current system performance, the impact of testing on component and system reliability, and how PSA considerations can be integrated into a comprehensive Maintenance, Surveillance, and Inspection (MSI) strategy. Although the system reliability requirements are an important part of the strategy evaluation, SAM techniques provide a systematic assessment within a broader context. Testing is only one part of an overall strategy focused on ensuring that component function is maintained through a combination of monitoring technologies (including testing), predictive techniques, and intrusive maintenance strategies. Each strategy is targeted to known component degradation mechanisms. This thinking can be extended to safety and safety related systems in general. Over the past 6 years, AECL has been working with CANDU utilities in the development and implementation of a comprehensive and integrated Plant Life Management (PLiM) program. As part of developing a comprehensive plant asset management approach, SAM techniques are used to develop a technical basis that not only works towards ensuring reliable operation of plant systems, but also facilitates the optimization and

  8. Safety-Related Contractor Activities at Nuclear Power Plants. New Challenges for Regulatory Oversight

    Chockie, Alan [Chockie Group International, Inc., Seattle, WA (United States)

    2005-09-15

    The use of contractors has been an integral and important part of the design, construction, operation, and maintenance of nuclear power plants. To ensure the safe and efficient completion of contracted tasks, each nuclear plant licensee has developed and refined formal contract management processes to meet their specific needs and plant requirements. Although these contract management processes have proven to be effective tools for the procurement of support and components tailored to the needs of nuclear power plants, contractor-related incidents and accidents have revealed some serious weaknesses with the implementation of these processes. Identifying and addressing implementation problems are becoming more complicated due to organizational and personnel changes affecting the nuclear power industry. The ability of regulators and licensees to effectively monitor and manage the safety-related performance of contractors will likely be affected by forthcoming organization and personnel changes due to: the aging of the workforce; the decline of the nuclear industry; and the deregulation of nuclear power. The objective of this report is to provide a review of current and potential future challenges facing safety-related contractor activities at nuclear power plants. The purpose is to assist SKI in establishing a strategy for the proactive oversight of contractor safety-related activities at Swedish nuclear power plants and facilities. The nature and role of contractors at nuclear plants is briefly reviewed in the first section of the report. The second section describes the essential elements of the contract management process. Although organizations have had decades of experience with the a contract management process, there remain a number of common implantation weaknesses that have lead to serious contractor-related incidents and accidents. These implementation weaknesses are summarized in the third section. The fourth section of the report highlights the

  9. Safety-Related Contractor Activities at Nuclear Power Plants. New Challenges for Regulatory Oversight

    Chockie, Alan

    2005-09-01

    The use of contractors has been an integral and important part of the design, construction, operation, and maintenance of nuclear power plants. To ensure the safe and efficient completion of contracted tasks, each nuclear plant licensee has developed and refined formal contract management processes to meet their specific needs and plant requirements. Although these contract management processes have proven to be effective tools for the procurement of support and components tailored to the needs of nuclear power plants, contractor-related incidents and accidents have revealed some serious weaknesses with the implementation of these processes. Identifying and addressing implementation problems are becoming more complicated due to organizational and personnel changes affecting the nuclear power industry. The ability of regulators and licensees to effectively monitor and manage the safety-related performance of contractors will likely be affected by forthcoming organization and personnel changes due to: the aging of the workforce; the decline of the nuclear industry; and the deregulation of nuclear power. The objective of this report is to provide a review of current and potential future challenges facing safety-related contractor activities at nuclear power plants. The purpose is to assist SKI in establishing a strategy for the proactive oversight of contractor safety-related activities at Swedish nuclear power plants and facilities. The nature and role of contractors at nuclear plants is briefly reviewed in the first section of the report. The second section describes the essential elements of the contract management process. Although organizations have had decades of experience with the a contract management process, there remain a number of common implantation weaknesses that have lead to serious contractor-related incidents and accidents. These implementation weaknesses are summarized in the third section. The fourth section of the report highlights the

  10. Study on some safety-related aspects of tyre use

    Jansen, S.T.H.; Schmeitz, A.J.C.; Maas, S.; Rodarius, C.; Akkermans, L.

    2014-01-01

    The tyre is a key component that affects road safety. The European commission has posted a tender aimed to study what measures on a European level can be taken in relation to the use of tyres to improve road safety. The results of this study, supported by a cost benefit analyses and carried out by

  11. The Burden of Peristomal Skin Complications on an Ostomy Population as Assessed by Health Utility and the Physical Component Summary of the SF-36v2®.

    Nichols, Thom R; Inglese, Gary W

    2018-01-01

    Body-altering surgery may affect perceptions of one's self. For those with abdominal stoma surgeries, altered perceptions amplified by peristomal skin condition can increase health burdens. To assess health utility and health-related quality of life in an adult US ostomy sample in the presence of three levels of peristomal skin condition: intact, moderately compromised, and severely compromised. The short form 36 health survey version 2, a generic health survey incorporating the six-dimensional health state short form preference-based utility index, was chosen to assess the sample. Analysis of covariance adjusted for age and time from surgery was used. The six-dimensional health state short form utilities for those with intact skin and physical component summary (PCS) levels indicating no physical limitations varied significantly from those with severely compromised skin and indicating the greatest degree of physical limitation (0.833 vs. 0.527). Peristomal skin condition decreases were associated with health utility decreases across all levels of the PCS. Because peristomal skin conditions are intermittent, the analysis presents quality-adjusted life-days (QALDs) per month. Ostomates with intact skin and PCS levels indicating no physical limitations demonstrated significant differences from those with severe skin condition and indicating the greatest degree of physical limitations (26.5 d/mo vs. 15.8 d/mo). As peristomal skin condition worsened, QALDs decreased across all levels of the PCS. A minimally important expected value of health was estimated to be an increase of 2.18 QALDs/mo. Successful treatment from a clinical perspective is more than the elimination of conditions-it is also a return of quality time to an individual. Copyright © 2018 International Society for Pharmacoeconomics and Outcomes Research (ISPOR). Published by Elsevier Inc. All rights reserved.

  12. 78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

    2013-05-01

    ... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for... regulatory guide (DG), DG-1235, ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants... entitled ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants'' is temporarily...

  13. Guidelines for evaluation of anchorage adequacy for safety-related equipment typically used on WWER-type NPPs

    Masopust, R.

    1999-01-01

    This report describes the criteria which should be met when the capacity evaluation of anchorage of safety related equipment is performed for the WWER type NPPs. It should be noted that these criteria were developed specifically for anchorage of WWER type equipment and components to the concrete or steel building structures and they cover different types of anchor bolts and other anchorage details which are typical just for the existing, constructed or reconstructed WWER type NPPs. The screening approach for verifying of equipment anchorage presented in this report is based on a combination of inspections, calculations, and engineering judgement

  14. The LBB methodology application results performed on the safety related piping of NPP V-1 in Jaslovske Bohunice

    Kupca, L.; Beno, P. [Nuclear Power Plants Research Institute, Trnava (Slovakia)

    1997-04-01

    A broad overview of the leak before break (LBB) application to the Slovakian V-1 nuclear power plant is presented in the paper. LBB was applied to the primary cooling circuit and surge lines of both WWER 440 type units, and also used to assess the integrity of safety related piping in the feed water and main steam systems. Experiments and calculations performed included analyses of stresses, material mechanical properties, corrosion, fatigue damage, stability of heavy component supports, water hammer, and leak rates. A list of analysis results and recommendations are included in the paper.

  15. Executive summary

    2009-01-01

    of lessons learned from other licensees, both nationally and internationally, and from other industries. Such a 'learning' organisation should be open-minded and have a fair blame culture in order to encourage recognising and reporting of both near misses and serious incidents. The use of relevant performance indicators to identify and address safety issues is also essential. Another set of attributes that was identified by the workshop participants concerned specific licensee and plant management actions. The management should have a systems perspective of their operations. This involves recognising and managing the safety-related interactions and relationships among various aspects of plant operations, including man-technology organisational interactions. Also, safety-related strategic thinking is an important attribute of a 'good' organisation. This requires taking both a medium and long-term perspective of the activities and developing robust and resilient strategies to ensure safe operations. Safety-oriented decision-making should be a routine part of the management activities with the resulting decisions effectively communicated both vertically and horizontally within the organisation. It was noted that there should be effective leadership to motivate the plant personnel to continually assess the safety implications of their situation and actions and communicate any concerns. This would support the ability of the organisation to recognise early warning signs, take appropriate action and to follow-up and assess the effectiveness of their actions. Other attributes of 'good' organisations include the development of clear lines of responsibility and accountability, the effective use of teams, and the establishing an 'intelligent customer' approach to contracted support. At the conclusions of the workshop, the participants distilled a number of key messages arising from the discussions and presentations. These can be grouped into the following categories: the

  16. Aging and service wear of air-operated valves used in safety-related systems at nuclear power plants

    Cox, D.F.; McElhaney, K.L.; Staunton, R.H.

    1995-05-01

    Air-operated valves (AOVs) are used in a variety of safety-related applications at nuclear power plants. They are often used where rapid stroke times are required or precise control of the valve obturator is required. They can be designed to operate automatically upon loss of power, which is often desirable when selecting components for response to design basis conditions. The purpose of this report is to examine the reported failures of AOVs and determine whether there are identifiable trends in the failures related to predictable causes. This report examines the specific components that comprise a typical AOV, how those components fail, when they fail, and how such failures are discovered. It also examines whether current testing frequencies and methods are effective in predicting such failures

  17. A Performance Improvement of Power Supply Module for Safety-related Controller

    Kim, Jong-Kyun; Yun, Dong-Hwa; Hwang, Sung-Jae; Lee, Myeong-Kyun; Yoo, Kwan-Woo

    2015-01-01

    In this paper, in relation to voltage shortage state when power supply module is a slave mode, the performance improvement by modifying a PFC(Power Factor Correction) circuit is presented. With the modification of the PFC circuit, the performance improvement in respect of the voltage shortage state when the power supply module is a slave mode is checked. As a result, POSAFE-Q PLC can ensure the stability with the redundant power supply module. The purpose of this paper is to improve the redundant performance of power supply module(NSPS-2Q). It is one of components in POSAFE-Q which is a PLC(Programmable Logic Controller) that has been developed for the evaluation of safety-related. Power supply module provides a stable power in order that POSAFE-Q can be operated normally. It is possible to be mounted two power supply modules in POSAFE-Q for a redundant(Master/Slave) function. So that even if a problem occurs in one power supply module, another power supply module will provide a power to POSAFE-Q stably

  18. An initial examination of aging related degradation in turbine drives and governors for safety related pumps

    Cox, D.F.

    1991-01-01

    This study is being performed to examine the relationship between time dependent degradation, and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized water reactor (PWR) plants, and the Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI) systems for Boiling Water Reactor (BWR) facilities. This research has been conducted by examining current information in NPRDS, reviewing Licensee Event Reports, and thoroughly investigating contacts with operating plant personnel, and by personal observation. The reported information was reviewed to determine the cause of the event and the method of discovery. From this data attempts have been made at determining the predictability of events and possible preventive measures that may be implemented. Findings in a recent study on the Auxiliary Feedwater System (NUREG/CR-5404) indicate that the turbine drive is the single largest contributor to AFW system degradation. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design seem to indicate that this equipment can be a reliable component in safety systems

  19. Report on safety related occurrences and reactor trips January 1 - June 30, 1985

    1986-01-01

    This is a systematically arranged report on all safety-related occurrences and reacotr trips in Swedish nuclear power plants in operation during the period from January 1 to June 30 1985. It is based on the reports submitted by the utilities to the Swedish Nuclear power Inspectorate according to Technical Specifications. Twice a year since 1974 the Inspectorate has issued a compilation on such reported occurrences and reactor trips. Starting with the compilation of the second half of 1982 some new features have been introduced. The most important change is that the volume of information has been increased. The full test, provided by the utilities when reporting the incidents, is now attached to the codified information and also the layout has been altered to facilitate reading. As in the previous reports the occurrences and reactor trips are arranged both alphabetically by facility name and chronologically by report number for each facility. Electricity generation charts for each facility are also presented. The primary purpose of this report is thus to present all the information furnished by utlities when they submit their reports according the Technical Specifications. The only evaluation made by the Inspecotrate is the categorization on the incidents. Like the previous reports this one also presents frequency of incidents as related to affected component, cause of incident etc. The difference is that only information reported by the utilities is used. This is the reason why a considerable proportion of the incidents are categorized as 'other fault'. (author)

  20. A Performance Improvement of Power Supply Module for Safety-related Controller

    Kim, Jong-Kyun; Yun, Dong-Hwa; Hwang, Sung-Jae; Lee, Myeong-Kyun; Yoo, Kwan-Woo [PONUTech Co., Seoul (Korea, Republic of)

    2015-10-15

    In this paper, in relation to voltage shortage state when power supply module is a slave mode, the performance improvement by modifying a PFC(Power Factor Correction) circuit is presented. With the modification of the PFC circuit, the performance improvement in respect of the voltage shortage state when the power supply module is a slave mode is checked. As a result, POSAFE-Q PLC can ensure the stability with the redundant power supply module. The purpose of this paper is to improve the redundant performance of power supply module(NSPS-2Q). It is one of components in POSAFE-Q which is a PLC(Programmable Logic Controller) that has been developed for the evaluation of safety-related. Power supply module provides a stable power in order that POSAFE-Q can be operated normally. It is possible to be mounted two power supply modules in POSAFE-Q for a redundant(Master/Slave) function. So that even if a problem occurs in one power supply module, another power supply module will provide a power to POSAFE-Q stably.

  1. Report on safety related occurrences and reactor trips January 1 - June 30, 1984

    1984-01-01

    This is a systematically arranged report on all safety-related occurrences and reactor trips in Swedish nuclear power plants in operation during the period from January 1 to June 30 1984. It is based on the reports submitted by the utilities to the Swedish Nuclear Inspectorate according to Technical Specifications. Twice a year since 1974 the Inspectorate has issued a compilation on such reported occurrences and reactor trips. Starting with the compilation of the second half of 1982 some new features have been introduced. The most important change is that the volume of information has been increased. The full text, provided by the utilities when reporting the incidents, is now attached to the codified information and also the layout has been altered to facilitate reading. As in the previous reports the occurrences and reactor trips are arranged both alphabetically by facility name and chronologically by report number for each facility. Electricity generation charts for each facility are also presented. The primary purpose of this report is thus to present all the information furnished by the utilities when they submit their reports according to Technical Specifications. The only evaluation made by the Inspectorate is the categorization on the incidents. Like the previous reports this one also presents frequency of incidents as related to affected component, cause of incident etc. The difference is that only information reported by the utilities is used. This is the reason why a considerable proportion of the incidents are categorized as other component or other fault. Sometime in the future, however, the Inspectorate plants to put out a special report containing its own analyses of the most interesting events along with processed statistics and other information. (author)

  2. Equivalent linear and nonlinear site response analysis for design and risk assessment of safety-related nuclear structures

    Bolisetti, Chandrakanth; Whittaker, Andrew S.; Mason, H. Benjamin; Almufti, Ibrahim; Willford, Michael

    2014-01-01

    Highlights: • Performed equivalent linear and nonlinear site response analyses using industry-standard numerical programs. • Considered a wide range of sites and input ground motions. • Noted the practical issues encountered while using these programs. • Examined differences between the responses calculated from different programs. • Results of biaxial and uniaxial analyses are compared. - Abstract: Site response analysis is a precursor to soil-structure interaction analysis, which is an essential component in the seismic analysis of safety-related nuclear structures. Output from site response analysis provides input to soil-structure interaction analysis. Current practice in calculating site response for safety-related nuclear applications mainly involves the equivalent linear method in the frequency-domain. Nonlinear time-domain methods are used by some for the assessment of buildings, bridges and petrochemical facilities. Several commercial programs have been developed for site response analysis but none of them have been formally validated for large strains and high frequencies, which are crucial for the performance assessment of safety-related nuclear structures. This study sheds light on the applicability of some industry-standard equivalent linear (SHAKE) and nonlinear (DEEPSOIL and LS-DYNA) programs across a broad range of frequencies, earthquake shaking intensities, and sites ranging from stiff sand to hard rock, all with a focus on application to safety-related nuclear structures. Results show that the equivalent linear method is unable to reproduce the high frequency acceleration response, resulting in almost constant spectral accelerations in the short period range. Analysis using LS-DYNA occasionally results in some unrealistic high frequency acceleration ‘noise’, which can be removed by smoothing the piece-wise linear backbone curve. Analysis using DEEPSOIL results in abrupt variations in the peak strains of consecutive soil layers

  3. Mergeable summaries

    Agarwal, Pankaj K.; Graham, Graham; Huang, Zengfeng

    2013-01-01

    We study the mergeability of data summaries. Informally speaking, mergeability requires that, given two summaries on two datasets, there is a way to merge the two summaries into a single summary on the two datasets combined together, while preserving the error and size guarantees. This property m...

  4. Commercial grade item (CGI) dedication of generators for nuclear safety related applications

    Das, R.K.; Hajos, L.G.

    1993-01-01

    The number of nuclear safety related equipment suppliers and the availability of spare and replacement parts designed specifically for nuclear safety related application are shrinking rapidly. These have made it necessary for utilities to apply commercial grade spare and replacement parts in nuclear safety related applications after implementing proper acceptance and dedication process to verify that such items conform with the requirements of their use in nuclear safety related application. The general guidelines for the commercial grade item (CGI) acceptance and dedication are provided in US Nuclear Regulatory Commission (NRC) Generic Letters and Electric Power Research Institute (EPRI) Report NP-5652, Guideline for the Utilization of Commercial Grade Items in Nuclear Safety Related Applications. This paper presents an application of these generic guidelines for procurement, acceptance, and dedication of a commercial grade generator for use as a standby generator at Salem Generating Station Units 1 and 2. The paper identifies the critical characteristics of the generator which once verified, will provide reasonable assurance that the generator will perform its intended safety function. The paper also delineates the method of verification of the critical characteristics through tests and provide acceptance criteria for the test results. The methodology presented in this paper may be used as specific guidelines for reliable and cost effective procurement and dedication of commercial grade generators for use as standby generators at nuclear power plants

  5. Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980

    McCormack, K.E.; Gallaher, R.B.

    1982-03-01

    This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).

  6. Commercial-grade motors in safety-related applications: Final report

    Holzman, P.M.

    1988-04-01

    The objective of this project was to discuss the process necessary to utilize commercial grade equipment in safety related applications and to provide utilities with guidance for accepting commercial grade motors for safety-related applications. The generic commercial-grade concepts presented in this report can be successfully applied to motors. Commercial grade item utilization has the greatest applicability to motors in ''mild'' environments, because these motors are essentially similar to commercial grade motors in materials, construction methods, and capabilities. The acceptance process is less applicable to motors that are subject to ''harsh'' environments during postulated accidents, because of the unique design features and testing required to qualify these motors

  7. Impact of pre-conditioning on the qualification of safety-related equipment

    Isgro, J.R.

    1982-01-01

    This paper shares some recent experiences on the effects of preconditioning on the qualification of safety-related equipment not located in a harsh environment. Environmental and seismic qualification testing programs were conducted following the guidelines of IEEE 323-1974, IEEE 344-1975 and appropriate IEEE daughter standards, where available. The examples that follow will illustrate the degree of pre-conditioning of safety-related equipment qualified to the requirements of IEEE-323-1974, and its effect on the outcome of the qualification program

  8. Management of the aging of critical safety-related concrete structures in light-water reactor plants

    Naus, D.J.; Oland, C.B.; Arndt, E.G.

    1990-01-01

    The Structural Aging Program has the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant safety-related structures for continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued-service determinations. Objectives, accomplishments, and planned activities under each of these tasks are presented. Major program accomplishments include development of a materials property data base for structural materials as well as an aging assessment methodology for concrete structures in nuclear power plants. Furthermore, a review and assessment of inservice inspection techniques for concrete materials and structures has been complete, and work on development of a methodology which can be used for performing current as well as reliability-based future condition assessment of concrete structures is well under way. 43 refs., 3 tabs

  9. Resistance ability evaluation of safety-related structures for the simulated aircraft accident

    Kim, Young Jin; Kim, Sung Woon; Choi, Jang Kyu [Daewoo E and C Co., Ltd., Suwon (Korea, Republic of)] (and others)

    2003-03-15

    Aircraft accidents on nuclear safety-related structures can cause severe damage to the safety of NPP(Nuclear Power Plant)s. To assess the safety of nuclear safety-related structures, the local damage and the dynamic response of global structures should be investigated together. This study have compared several local damage assessment formulas suggested for aircraft as an impactor, and have set the assessment system of local damage for impact-proof design of NPP containment buildings. And the local damage of nuclear safety-related structures in operation in Korea for commercial aircraft as impactor have been estimated. Impact load-time functions of the aircraft crash have been decided to assessment the safety of nuclear safety-related structures against the intentional colliding of commercial aircraft. Boeing 747 and Boeing 767 is selected as target aircraft based on the operation frequencies and weights. Comparison of the fire analysis methods showed that the method considering heat convection and radiation is adequate for the temperature analysis of the aircraft fuel fire. Finally, the study covered the analysis of the major structural drawings and design drawings with which three-dimensional finite element model analysis is expected to be performed.

  10. 78 FR 67206 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

    2013-11-08

    ... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Revision to regulatory guide; issuance..., ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants.'' This RG is being revised to provide... Operators Installed Inside the Containment of Nuclear Power Plants,'' dated January 1974. ADDRESSES: Please...

  11. Symposium summary

    Wall, G.

    1990-01-01

    A summary is provided of the issues discussed at the climate change implications for water and ecological resources conference, and recommendations that came out of the conference. The objectives of the meeting were to present and discuss results of recent climate change experiments undertaken in Canada; evaluate a variety of climate models and impact analyses and to develop methods and strategies for future study; and to establish working linkages between modellers and analysts in the fields of climate, hydrology, and ecosystem research, as well as between social scientists and policy makers interested in the implications of climate change. Recommendations were made in the five areas of research, monitoring, risk assessment, policy and information dissemination. Additional research should be undertaken to foster improved understanding of relationships between climate, climate change, and ecological and human processes. A suitable monitoring program, including a national wetlands monitoring program, should be established. Risk assessments should be undertaken to evaluate vulnerabilities of ecosystem components, to assess options, and to provide the information required to develop and implement appropriate policy objectives. The impacts of a range of public policy responses and feedbacks should be assessed. The dissemination of well-targeted and accurate information is vital if basic societal attitudes regarding the value of water and ecosystems are to be changed

  12. New trends in the evaluation and implementation of the safety-related operating experience associated with NRC-licensed reactors

    Michelson, C.; Heltemes, C.J.

    1981-01-01

    This article is an overview of the Nuclear Regulatory Commission program for the evaluation and dissemination of the safety-related operating experience associated with all NRC-licensed reactors. It discusses the historical background and past problems that led to the recent formation of NRC's Office for Analysis and Evaluation of Operational Data (AEOD) and details its activities, organization, staffing, and proposed analysis and evaluation methodology. The programs of industry organizations and nuclear plant licensees and the integration of foreign operating experience are included in the overview. The problems and limitations of the Licensee Event Report (LER) program and the Nuclear Plant Reliability Data system program are discussed. The AEOD analysis and evaluation methodology program includes some new improvements in the assessment of safety-related operating experience. Of particular note is the sequence coding and search procedure being developed by AEOD under a contract with the Nuclear Safety Information Center at the Oak Ridge National Laboratory. This computer-based retrieval system will have markedly improved search strategy capability for such items as commoncause failures or complex system interactions involving various failure sequences and other relationships associated with an event. The system retrieves failure data and information on the principal LER occurrence and on related component and system responses. The computer-generated Power Reactor Watch List enables AEOD to monitor all critical or unusual situations warranting close attention because of potential public health and safety. This listing is supported by a preestablished computer search strategy of the historical data base permitting identification of all past events and statistical information that are applicable to the situation being watched

  13. Research Summaries

    Brock, Stephen E., Ed.

    2011-01-01

    This article presents summaries of three articles relevant to school crisis response: (1) "Factors Contributing to Posttraumatic Growth," summarized by Steve DeBlois; (2) "Psychological Debriefing in Cross-Cultural Contexts" (Stacey Rice); and (3) "Brain Abnormalities in PTSD" (Sunny Windingstad). The first summary reports the findings of a…

  14. Dedication for Safety-Related Fuses used in Class-1E Power System

    Hong, Younghee

    2014-01-01

    The safety-related fuses used in class-1E power system provide overcurrent protection for electrical system and isolate the class 1E circuit from a fault or overload condition. These days, the number of nuclear grade suppliers has been reduced. Accordingly, commercial grade, instead of safety-related, fuses are procured and used in the utilities through the dedication process. Therefore, this paper introduces the commercial grade fuse dedication process/engineering and how to assure the quality requirements with this process and engineering. The fuses used in class-1E power system are to protect overcurrent and to isolate fault. Therefore the fuse for acceptance in order to improve the quality and reliability for commercial grade fuses shall be dedicated. The fuse resistance value may be useful as an indicator of acceptance. The current carrying capacity test can change the fuse performance properties. Therefore these critical characteristics are needed for additional review and analysis with fuse manufactures

  15. Recent progress in safety-related applications of reactor noise analysis

    Hirota, Jitsuya; Shinohara, Yoshikuni; Saito, Keiichi

    1982-01-01

    Recent progress in safety-related applications of reactor noise analysis is reviewed, mainly referring to various papers presented at the Third Specialists' Meeting on Reactor Noise (SMORN-III) held in Tokyo in 1981. Advances in application of autoregressive model, coherence analysis and pattern recognition technique are significant since SMORN-II in 1977. Development of reactor diagnosis systems based on noise analysis is in progress. Practical experiences in the safety-related applications to power plants are being accumulated. Advances in quantitative monitoring of vibration of internal structures in PWR and diagnosis of core stability and control system characteristics in BWR are notable. Acoustic methods are also improved to detect sodium boiling in LMFBR. The Reactor Noise Analysis Benchmark Test performed by Japan in connection with SMORN-III is successful so that it is possible to proceed to the second stage of the benchmark test. (author)

  16. Synergistic behaviour of nuclear radiation, temperature-humidity extremes and LOCA situation on safety and safety-related equipment in Indian nuclear power plants

    Kulkarni, R.D.; Bora, J.S.; Prakash, Ravi; Agarwal, Vivek; Sundersingh, V.P.

    2002-01-01

    Full text: The general philosophy for the instrumentation in nuclear power plants is based on the use of equipment/instruments which are capable of continuous satisfactory operation over a long period of time with minimum attention. Long term reliability under varying service conditions is of prime importance. The reliability of nuclear power plant depends on the reliability of safety and safety-related electronic instruments/ equipment used for performing the crucial tasks. The electrical and electronic systems/ circuits/ components of the equipment used in reactor safety systems (e.g. reactor protection system, emergency core cooling system, etc.) and reactor safety-related systems (e.g. reactor containment isolation and cooling system, reactor shutdown system, etc.) are responsible for safe and reliable operation of a nuclear power plant. The performance of reactor safety and safety-related equipment/instruments viz. pressure and differential pressure transmitter, amplifier for ion chamber, etc. has been evaluated under synergistic atmosphere including LOCA to find out the critical link in the circuits and subsequent modifications are suggested. The mathematical representation of the generated database has been done to estimate the life span of the instruments and accordingly the guidelines has been prepared for the operational staff to avoid the forced outage of the plant. All the details are included and mathematical models are presented to predict the future performances

  17. Seismic analysis of the safety related piping and PCLS of the WWER-440 NPP

    Berkovski, A.M.; Kostarev, V.V.; Schukin, A.J.; Boiadjiev, Z.; Kostov, M.

    2001-01-01

    This paper presents the results of seismic analysis of Safety Related Piping Systems of the typical WWER-440 NPP. The methodology of this analysis is based on WANO Terms of Reference and ASME BPVC. The different possibilities for seismic upgrading of Primary Coolant Loop System (PCLS) were considered. The first one is increasing of hydraulic snubber units and the second way is installation of limited number of High Viscous Dampers (HVD). (author)

  18. Application of project management methodology in design management of nuclear safety related structure

    Chen Mao

    2004-01-01

    This paper focuses on the application of project management methodology in the design management of Nuclear Safety Related Structure (NSRS), considering the design management features of its civil construction. Based on the experiences from the management of several projects, the project management triangle is proposed to be used in the management, to well treat the position of design interface in the project management. Some other management methods are also proposed

  19. Emergency Diesel: Safety-related instrumentation and control with programmable logic controllers

    Breidenich, G.; Luedtke, M.

    2004-01-01

    This report presents a new concept for the design of emergency diesel equipment protection circuits as a part of the safety related instrumentation in the nuclear power plant Biblis, units A and B. The concept was implemented with state of the art SIMATIC S7/316 programmable logic controllers (PLCs) and can be adapted to any system with high availability requirements (e.g. power plant turbines, aircraft engines, mining pumps etc). (orig.)

  20. Peer training of safety-related skills to institutional staff: benefits for trainers and trainees.

    van Den Pol, R A; Reid, D H; Fuqua, R W

    1983-01-01

    A peer training program, in which experienced staff trained new staff, was evaluated as a method for teaching and maintaining safety-related caregiver skills in an institutional setting for the developmentally disabled. Three sets of safety-type skills were assessed in simulated emergency situations: responding to facility fires, managing aggressive attacks by residents, and assisting residents during convulsive seizures. Using a multiple-baseline research design, results indicated that the p...

  1. PWR composite materials use. A particular case of safety-related service water pipes

    Pays, M.F.; Le Courtois, T.

    1997-11-01

    This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during 'lifetime'); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author)

  2. Development of FPGA-based safety-related I and C systems

    Goto, Y.; Oda, N.; Miyazaki, T.; Hayashi, T.; Sato, T.; Igawa, S. [08, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); 1, Toshiba-cho, Fuchu, Tokyo 183-8511 (Japan)

    2006-07-01

    Toshiba has developed Non-rewritable (NRW) Field Programmable Gate Array (FPGA)-based safety-related Instrumentation and Control (I and C) system [1]. Considering application to safety-related systems, nonvolatile and non-rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. FPGA is a device which consists only of defined digital circuit: hardware, which performs defined processing. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing unit (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. The system which Toshiba developed this time is Power Range Monitor (PRM). Toshiba is planning to expand application of FPGA-based technology by adopting this development method to the other safety-related systems from now on. (authors)

  3. Development of FPGA-based safety-related instrumentation and control systems

    Oda, N.; Tanaka, A.; Izumi, M.; Tarumi, T.; Sato, T. [Toshiba Corporation, Isogo Nuclear Engineering Center, Yokohama (Japan)

    2004-07-01

    Toshiba has developed systems which perform signal processing by field programmable gate arrays (FPGA) for safety-related instrumentation and control systems. FPGA is a device which consists only of defined digital circuit: hardware, which performs defined processing. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing units (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. Considering application to safety-related systems, nonvolatile and non rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. The systems which Toshiba developed this time are Power range Monitor (PRM) and Trip Module (TM). These systems are compatible with the conventional analog-based systems and the CPU-based systems. Therefore, requested cost for upgrading will be minimized. Toshiba is planning to expand application of FPGA-based technology by adopting this development method to the other safety-related systems from now on. (authors)

  4. PWR composite materials use. A particular case of safety-related service water pipes

    Pays, M.F.; Le Courtois, T

    1997-11-01

    This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during `lifetime`); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author) 2 refs.

  5. Guards: An approach safety-related systems using cots example of MMI and reactor automation in nuclear submarine application

    Brun, M.

    1998-01-01

    For at least 10 years, the nuclear industry designs and licences specific digital safety-critical systems (IEC 1226 class A). One key issue for future programs is to design and licence safety-related systems providing more complex functions and using Commercial-Off-The-Shelf components. This issue is especially raised for Reactor automation and Man-Machine-Interface. The usual I and C (Instrumentation and Control) organisation for these functions is based on redundancy between a commercial, up-to-date, unclassified > system and a simplified classified > system using traditional technologies. It clearly appears that such organisation is not satisfying from the point of view of people who have actually to operate these systems: The operator is supposed not to trust the normal system and rely on the back-up system which is less helpful and that he use very few. This paper presents a new approach to that problem using COTS components in low-level layers, safety architecture and mechanisms at medium level layer (GUARDS architecture developed in the current ESPRIT project number 20716), and a pre-validated functional layer. The aim of this solution is to comply with the > IEC 1226 class B requirements, at lower overall cost (design, implementation, licensing, long term confidence). This approach is illustrated by its application in Man-Machine-Interface (MMI) for our future program of Nuclear submarine. (author)

  6. Meteorological Summaries

    National Oceanic and Atmospheric Administration, Department of Commerce — Multi-year summaries of one or more meteorological elements at a station or in a state. Primarily includes Form 1078, a United States Weather Bureau form designed...

  7. Survey Summary

    U.S. Department of Health & Human Services — Nursing home summary information for the Health and Fire Safety Inspections currently listed on Nursing Home Compare, including dates of the three most recent...

  8. Using naturalistic driving data to explore the association between traffic safety-related events and crash risk at driver level.

    Wu, Kun-Feng; Aguero-Valverde, Jonathan; Jovanis, Paul P

    2014-11-01

    There has been considerable research conducted over the last 40 years using traffic safety-related events to support road safety analyses. Dating back to traffic conflict studies from the 1960s these observational studies of driver behavior have been criticized due to: poor quality data; lack of available and useful exposure measures linked to the observations; the incomparability of self-reported safety-related events; and, the difficulty in assessing culpability for safety-related events. This study seeks to explore the relationships between driver characteristics and traffic safety-related events, and between traffic safety-related events and crash involvement while mitigating some of those limitations. The Virginia Tech Transportation Institute 100-Car Naturalistic Driving Study dataset, in which the participants' vehicles were instrumented with various cameras and sensors during the study period, was used for this study. The study data set includes 90 drivers observed for 12-13 months driving. This study focuses on single vehicle run-off-road safety-related events only, including 14 crashes and 182 safety-related events (30 near crashes, and 152 crash-relevant incidents). Among the findings are: (1) drivers under age 25 are significantly more likely to be involved in safety-related events and crashes; and (2) significantly positive correlations exist between crashes, near crashes, and crash-relevant incidents. Although there is still much to learn about the factors affecting the positive correlation between safety-related events and crashes, a Bayesian multivariate Poisson log-normal model is shown to be useful to quantify the associations between safety-related events and crash risk while controlling for driver characteristics. Copyright © 2014 Elsevier Ltd. All rights reserved.

  9. Seismic qualification of multiple interconnected safety-related cabinets in a high seismic zone

    Khan, M.R.; Chen, W.H.W.; Wang, T.Y.

    1993-01-01

    Certain safety-related multiple, interconnected electrical cabinets and the devices contained therein are required to perform their intended safety functions during and after a design basis seismic event. In general, seismic testing is performed to ensure the structural integrity of the cabinets and the functionality of their associated devices. Constrained by the shake table capacity, seismic testing is usually performed only for a limited number of interconnected cabinets. Also, original shake table tests performed usually did not provide detailed response information at various locations inside the cabinets. For operational and maintenance purposes, doors and panels of some cabinets may need to be opened while the adjacent cabinets are required to remain functional. In addition, in-cabinet response spectra need to be generated for the seismic qualification of new devices and the replacement parts. Consequently, seismic analysis of safety-related multiple, interconnected cabinets is frequently required for configurations which are different from the original tested conditions. This paper presents results of seismic tests of three interconnected safety-related cabinets and finite element analyses performed to compare the analytical results with those obtained from the cabinet seismic tests. Parametric analyses are performed to determine how many panels and doors can be opened while the adjacent cabinets still remain functional. The study indicates that for cabinets located in a high seismic zone, the critical damping of the cabinet is significantly higher than 5% to 7% typically used in qualifying electrical equipment. For devices mounted on the cabinet doors to performed their intended safety function, it requires stiffening of doors and that these doors be properly bolted to the cabinet frame. It also shows that even though doors and panels bolted to the cabinet frame are the primary seismic resistant element of the cabinet, opening of a limited number of them

  10. Risk-based evaluation tool for safety-related maintenance involving scaffolding

    Stevens, C.; Azizi, M.; Massman, M.

    1988-01-01

    The US Nuclear Regulatory Commission (NRC) has expressed a general concern that transient materials in and around safety systems at nuclear power plants represent a seismic safety hazard to the plant, in particular, the uncontrolled use of scaffolding during maintenance activities. Currently, most plants perform a seismic safety analysis for all uses of scaffolding near safety-related equipment to determine appropriate tie-down locations, scaffolding reinforcements, etc. This is both time-consuming and, for the most part, unnecessary. A workable engineering solution based on risk analysis techniques has been developed and is being used at the Palo Verde nuclear generating station (PVNGS)

  11. Safety-related instrumentation and control systems for nuclear power plants

    1984-01-01

    This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety but are not safety systems. The Guide is intended to expand paragraphs 3.1, 3.2 and 3.3 of the Code of Practice on Design for Safety of Nuclear Power Plants (IAEA Safety Series No.50-C-D) in the area of I and C systems important to safety and refers to them as safety-related I and C systems. It also gives guidance and enumerates requirements for multiplexing and the use of the digital computers employed in this area

  12. Staff report on the environmental qualification of safety-related electrical equipment

    1977-12-01

    The current NRC safety review process for nuclear power plants includes criteria related to the qualification of certain electrical equipment. These criteria require that electrical equipment important to safety must be qualified to function in the environment that might result from various accident conditions. Although such criteria have been applied since the early days of commercial nuclear power, the details of these criteria have been changed over the years. The evolution of environmental qualification of safety-related electrical equipment is described in Appendix A

  13. Modification and backfitting at the Oskarshamn Nuclear Power Plant Unit 2 in safety related systems

    Karlsson, Leif; Nilsson, Ove; Lidh, B.

    1995-05-01

    This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Oskarshamn-2, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 3 refs

  14. Aging and service wear of spring-loaded pressure relief valves used in safety-related systems at nuclear power plants

    Staunton, R.H.; Cox, D.F.

    1995-03-01

    Spring-loaded pressure relief valves (PRVS) are used in some safety-related applications at nuclear power plants. In general, they are used in systems where, during accidents, pressures may rise to levels where pressure safety relief is required for protection of personnel, system piping, and components. This report documents a study of PRV aging and considers the severity and causes of service wear and how it is discovered and corrected in various systems, valve sizes, etc. Provided in this report are results of the examination of the recorded failures and identification of trends and relationships/correlations in the failures when all failure-related parameters are considered. Components that comprise a typical PRV, how those components fail, when they fail, and the current testing frequencies and methods are also presented in detail

  15. Aging and service wear of spring-loaded pressure relief valves used in safety-related systems at nuclear power plants

    Staunton, R.H.; Cox, D.F. [Oak Ridge National Lab., TN (United States)

    1995-03-01

    Spring-loaded pressure relief valves (PRVS) are used in some safety-related applications at nuclear power plants. In general, they are used in systems where, during accidents, pressures may rise to levels where pressure safety relief is required for protection of personnel, system piping, and components. This report documents a study of PRV aging and considers the severity and causes of service wear and how it is discovered and corrected in various systems, valve sizes, etc. Provided in this report are results of the examination of the recorded failures and identification of trends and relationships/correlations in the failures when all failure-related parameters are considered. Components that comprise a typical PRV, how those components fail, when they fail, and the current testing frequencies and methods are also presented in detail.

  16. A Development of the Calibration Tool Applied on Analog I/O Modules for Safety-related Controller

    Kim, Jong-Kyun; Yun, Dong-Hwa; Lee, Myeong-Kyun; Yoo, Kwan-Woo

    2016-01-01

    The purpose of this paper is to develop the calibration tool for analog input/output(I/O) modules. Those modules are components in POSAFE-Q which is a programmable logic controller(PLC) that has been developed for the evaluation of safety-related. In this paper, performance improvement of analog I/O modules is presented by developing and applying the calibration tool for each channel in analog I/O modules. With this tool, the input signal to an analog input module and the output signal from an analog output module are able to be satisfied with a reference value of sensor type and an accuracy of all modules. With RS-232 communication, the manual calibration tool is developed for analog I/O modules of an existing and up-to-date version in POSAFE-Q PLC. As a result of applying this tool, the converted value is performant for a type of input sensor and an accuracy of analog I/O modules

  17. Performance Monitoring for Nuclear Safety Related Instrumentation at PUSPATI TRIGA Reactor (RTP)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2015-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on performance monitoring for nuclear safety related instrumentation in TRIGA PUSPATI Reactor (RTP) of based on various parameter of reactor safety instrument channel such as log power, linear power, Fuel temperature, coolant temperature will take into consideration. Methodology of performance on estimation and monitoring is to evaluate and analysis of reactor parameters which is important of reactor safety and control. And also to estimate power measurement, differential of log and linear power and fuel temperature during reactor start-up, operation and shutdown .This study also focus on neutron power fluctuation from fission chamber during reactor start-up and operation. This work will present result of performance monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that performance of nuclear safety related instrumentation will improved the reactor control and safety parameter during reactor start-up, operation and shutdown. (author)

  18. A PLC generic requirements and specification for safety-related applications in nuclear power plants

    Han, Jea Bok; Lee, C. K.; Lee, D. Y.

    2001-12-01

    This report presents the requirements and specification to be applied to the generic qualification of programmable Logic Controller(PLC), which is being developed as part of the KNICS project, 'Development of the Digital Reactor Safety Systems' of which purpose is the application to safety-related instrumentation and control systems in nuclear power plants. This report defines the essential and critical characteristics that shall be included as part of a PLC design for safety-related application. The characteristics include performance, reliability, accuracy, the overall response time from an input to the PLC exceeding it trip condition to the resulting outputs, and the specification of processors and memories in digital controller. It also specifies the quality assurance process for software development, dealing with executive software, firmware, application software tools for developing the application software, and human machine interface(HMI). In addition, this report reviews the published standards and guidelines that are required for the PLC development and the quality assurance processes such as environment requirements, seismic withstand requirements, EMI/RFI withstand requirements, and isolation test

  19. Commercial grade item (CGI) dedication of MDR relays for nuclear safety related applications

    Das, Ranjit K.; Julka, Anil; Modi, Govind

    1994-08-01

    MDR relays manufactured by Potter & Brumfield (P&B) have been used in various safety related applications in commercial nuclear power plants. These include emergency safety features (ESF) actuation systems, emergency core cooling systems (ECCS) actuation, and reactor protection systems. The MDR relays manufactured prior to May 1990 showed signs of generic failure due to corrosion and outgassing of coil varnish. P&B has made design changes to correct these problems in relays manufactured after May 1990. However, P&B does not manufacture the relays under any 10CFR50 Appendix B quality assurance (QA) program. They manufacture the relays under their commercial QA program and supply these as commercial grade items. This necessitates CGI Dedication of these relays for use in nuclear-safety-related applications. This paper presents a CGI dedication program that has been used to dedicate the MDR relays manufactured after been used to dedicate the MDR relays manufactured after May 1990. The program is in compliance with current Nuclear Regulatory Commission (NRC) and Electric Power Research Institute (EPRI) guidelines and applicable industry standards; it specifies the critical characteristics of the relays, provides the tests and analysis required to verify the critical characteristics, the acceptance criteria for the test results, performs source verification to quality P&B for its control of the critical characteristics, and provides documentation. The program provides reasonable assurance that the new MDR relays will perform their intended safety functions.

  20. A new approach to determine the environmental qualification requirements for the safety related equipment

    Hasnaoui, C.; Parent, G.

    2000-01-01

    The objective of the environmental qualification of safety related equipment is to ensure that the plant defense-in-depth is not compromised by common mode failures following design basis accidents with a harsh environment. A new approach based on safety functions has been developed to determine what safety-related equipment is required to function during and after a design basis accident, as well as their environmental qualification requirements. The main feature of this approach is to use auxiliary safety functions established from safety requirements as credited in the safety analyses. This approach is undertaken in three steps: identification of the auxiliary safety functions of each main safety function; determination of the main equipment groups required for each auxiliary safety function; and review of the safety analyses for design basis accidents in order to determine the credited auxiliary safety functions and their mission times for each accident scenario. Some of the benefits of the proposed approach for the determination of the safety environmental qualification requirements are: a systematic approach for the review of safety analyses based on a safety function check list, and the insurance, with the availability of the safety functions, that Gentilly-2 defense-in-depth would not be compromised by design basis accidents with a harsh environment. (author)

  1. A Study of Cyber Security Activities for Development of Safety-related Controller

    Lee, Myeongkyun; Song, Seunghwan; Yoo, Kwanwoo; Yun, Donghwa [Korea Univ., Seoul (Korea, Republic of)

    2014-05-15

    Nuclear Power Plant Regulatory guide describes the regulatory requirements to implement cyber security activities to ensure that design and operate to respond to cyber threats that exploited to vulnerability of digital-based technologies associated with safety-related digital instrumentation and control systems at nuclear power plants. Cyber security activities coverage is instrumentation and control systems to perform safety functions and digital-based equipment to use development, test, analysis and asset for instrumentation and control systems. Regulatory guidance is required to the cyber security activities that should be performed in each development phase of safety-related controller. Development organization should establish and implement to cyber security plans for responding to cyber threats throughout each lifecycle phase and the result of the cyber security activities should be generated to the documents. In addition, the independent verification and validation organization should perform simulated penetration test for enhancing response capabilities to cyber security threats and development organization should establish and implement response hardening solutions for the cyber security vulnerabilities identified in the simulated penetration test.

  2. A Study of Cyber Security Activities for Development of Safety-related Controller

    Lee, Myeongkyun; Song, Seunghwan; Yoo, Kwanwoo; Yun, Donghwa

    2014-01-01

    Nuclear Power Plant Regulatory guide describes the regulatory requirements to implement cyber security activities to ensure that design and operate to respond to cyber threats that exploited to vulnerability of digital-based technologies associated with safety-related digital instrumentation and control systems at nuclear power plants. Cyber security activities coverage is instrumentation and control systems to perform safety functions and digital-based equipment to use development, test, analysis and asset for instrumentation and control systems. Regulatory guidance is required to the cyber security activities that should be performed in each development phase of safety-related controller. Development organization should establish and implement to cyber security plans for responding to cyber threats throughout each lifecycle phase and the result of the cyber security activities should be generated to the documents. In addition, the independent verification and validation organization should perform simulated penetration test for enhancing response capabilities to cyber security threats and development organization should establish and implement response hardening solutions for the cyber security vulnerabilities identified in the simulated penetration test

  3. Logic qualification of FPGA-based safety-related I and C systems

    Hayashi, Toshifumi; Oda, Naotaka; Ito, Toshiaki; Miyazaki, Tadashi; Haren, Yasuhiko

    2009-01-01

    We established a logic qualification method for FPGA-Based I and C safety-related use in Nuclear Power Plants Systems. The FPGA is a programmable logic device and has advantages that the programming is rigorous, simple verifiable, and the technology is stable. However, logic qualification of FPGA had been an issue to be solved when it is used in the safety-related systems, because FPGA is relatively new technology for the nuclear power industry. We employed a software-life cycle approach, because its development process is similar to that of conventional computer-based systems. There are some differences between the FPGA-Based systems and the computer-based systems in the implementation and integration of logic. We examined the FPGA logic implementation and integration process to identify any FPGA-Based system specific hazards. The identified hazards are (1) small logic errors, (2) timing errors, (3) logic synthesis errors, (4) place and route errors, and (5) logic embedding errors. We took the appropriate countermeasures to mitigate these hazards, and employed this logic qualification method in the qualification of the Power Range Monitor System for BWR Power Plants. (author)

  4. Research Summaries

    Brock, Stephen E., Ed.

    2010-01-01

    This column features summaries of research articles from 3 recent crisis management publications. The first, "School Shootings and Counselor Leadership: Four Lessons from the Field" summarized by Kristi Fenning, was conducted as the result of the increased demand for trained crisis personnel on school campuses. Survey participants were…

  5. Conference summaries

    1986-01-01

    This volume contains conference summaries of the international conference on radioactive waste management of the Canadian Nuclear Society. Topics of discussion include: storage and disposal; hydrogeology and geochemistry; transportation; buffers and backfill; public attitudes; tailings; site investigations and geomechanics; concrete; economics; licensing; matrix materials and container design; durability of fuel; biosphere modelling; radioactive waste processing; and, future options

  6. Executive summary

    van Nimwegen, N.; van Nimwegen, N.; van der Erf, R.

    2009-01-01

    The Demography Monitor 2008 gives a concise overview of current demographic trends and related developments in education, the labour market and retirement for the European Union and some other countries. This executive summary highlights the major findings of the Demography Monitor 2008 and further

  7. Interim staff position on environmental qualification of safety-related electrical equipment: including staff responses to public comments. Regulatory report

    Szukiewicz, A.J.

    1981-07-01

    This document provides the NRC staff positions regarding selected areas of environmental qualification of safety-related electrical equipment, in the resolution of Unresolved Safety Issue A-24, 'Qualification of Class IE Safety-Related Equipment.' The positions herein are applicable to plants that are or will be in the construction permit (CP) or operating license (OL) review process and that are required to satisfy the requirements set forth in either the 1971 or the 1974 version of IEEE-323 standard

  8. Review of domestic and international experience on optimization of tests planning for safety related systems at NPP

    Skalozubov, V.I.; Komarov, Yu.A.; Kolykanov, V.N.; Kochneva, V.Yu.; Gablaya, T.V.

    2009-01-01

    There are represented the basic requirements of normative and operating documents on test periodicity of safety related systems at NPPs, sets out the theoretical methods of test optimization of the technical systems, and analyses foreign engineering methods for changing test periodicity of the NPP systems. Based on this review analyses further tasks are formulated for improvement of the methodical base of optimization of tests planning for safety related systems

  9. Using field feedback to estimate failure rates of safety-related systems

    Brissaud, Florent

    2017-01-01

    The IEC 61508 and IEC 61511 functional safety standards encourage the use of field feedback to estimate the failure rates of safety-related systems, which is preferred than generic data. In some cases (if “Route 2_H” is adopted for the 'hardware safety integrity constraints”), this is even a requirement. This paper presents how to estimate the failure rates from field feedback with confidence intervals, depending if the failures are detected on-line (called 'detected failures', e.g. by automatic diagnostic tests) or only revealed by proof tests (called 'undetected failures'). Examples show that for the same duration and number of failures observed, the estimated failure rates are basically higher for “undetected failures” because, in this case, the duration observed includes intervals of time where it is unknown that the elements have failed. This points out the need of using a proper approach for failure rates estimation, especially for failures that are not detected on-line. Then, this paper proposes an approach to use the estimated failure rates, with their uncertainties, for PFDavg and PFH assessment with upper confidence bounds, in accordance with IEC 61508 and IEC 61511 requirements. Examples finally show that the highest SIL that can be claimed for a safety function can be limited by the 90% upper confidence bound of PFDavg or PFH. The requirements of the IEC 61508 and IEC 61511 relating to the data collection and analysis should therefore be properly considered for the study of all safety-related systems. - Highlights: • This paper deals with requirements of the IEC 61508 and IEC 61511 for using field feedback to estimate failure rates of safety-related systems. • This paper presents how to estimate the failure rates from field feedback with confidence intervals for failures that are detected on-line. • This paper presents how to estimate the failure rates from field feedback with confidence intervals for failures that are only revealed by

  10. Group contribution modelling for the prediction of safety-related and environmental properties

    Frutiger, Jerome; Abildskov, Jens; Sin, Gürkan

    warming potential and ozone depletion potential. Process safety studies and environmental assessments rely on accurate property data. Safety data such as flammability limits, heat of combustion or auto ignition temperature play an important role in quantifying the risk of fire and explosions among others......We present a new set of property prediction models based on group contributions to predict major safety-related and environmental properties for organic compounds. The predicted list of properties includes lower and upper flammability limits, heat of combustion, auto ignition temperature, global...... models like group contribution (GC) models can estimate data. However, the estimation needs to be accurate, reliable and as little time-consuming as possible so that the models can be used on the fly. In this study the Marrero and Gani group contribution (MR GC) method has been used to develop the models...

  11. Safety-related site characteristics - a relative comparison of the Forsmark reference areas

    Winberg, Anders

    2010-12-01

    SKB has over the years from 2002 to 2008 conducted site investigations in Forsmark and Laxemar, with associated site modeling, design and safety analysis. In mid-2009 Forsmark was selected on the basis of analysis made as site for a future repository for spent nuclear fuel. Based on defined safety-related geoscientific location factors data from Forsmark are compared in relative terms with data from a number of locations in Sweden, previously studied by SKB. The factors compared include: the rock's composition and structures, future climate evolution, rock mechanical conditions, earthquakes, groundwater flow, groundwater composition, delay of solutes, and the ability to characterize and describe the location. Past comparisons of these properties for the selected sites show that none of these sites collectively show any significant benefit over Forsmark site for a repository. This does not preclude that there may be places on the basis of an overall assessment of geoscientific location factors could be equivalent to Forsmark

  12. Role of security during safety-related emergencies at nuclear power plants

    Cardwell, R.G.; Moul, D.A.; McBride, J.A.; Wilson, C.W.

    1984-03-01

    This report provides an analysis of the literature and on-site data gathering relating to the actions of security forces at licensed nuclear power plants during safety-related emergencies. Literature search findings and results of on-site data gathering are furnished and subjected to analysis. Taking into account the analysis provided, appropriate recommendations are presented. Recommendations are keyed as to how improvements can be made in the regulatory approach and licensee planning and procedures as they relate to the subject matter under examination. In addition, certain technological problems and issues are examined within the context of the study. Appendices provide the results of the literature search, an annotated bibliography, the Data Collection Guide used, and additional details regarding certain aspects of the study that are relevant for further explication of the body of the report

  13. Evaluation of Generic Issue 57: Effects of fire protection system actuation on safety-related equipment

    Lambright, J.; Bohn, M.; Lynch, J.; Ross, S.; Brosseau, D.

    1992-12-01

    Nuclear power plants have experienced actuations of fire protection systems (FPSs) under conditions for which these systems were not intended to actuate and also have experienced advertent actuations with the presence of a fire. These actuations have often damaged safety-related equipment. A review of the impact of past occurrences of both types of such events and their impact on plant safety systems, an analysis of the risk impacts of such events on nuclear power plant safety, and a cost-benefit analysis of potential corrective measures have been performed. Thirteen different scenarios leading to actuation of fire protection systems due to a variety of causes were identified. These scenarios ranged from inadvertent actuation caused by human error to hardware failure, and include seismic root causes and seismic/fire interactions. A quantification of these thirteen root causes, where applicable, was performed on generically applicable scenarios. This document, Volume 4, contains appendices E and F of this report

  14. Seismic fragility testing of naturally-aged, safety-related, class 1E battery cells

    Bonzon, L.L.; Hente, D.B.; Kukreti, B.M.; Schendel, J.S.; Black, D.A.; Paulsen, G.D.; Tulk, J.D.; Janis, W.J.; Aucoin, B.D.

    1984-01-01

    The concern over seismic susceptibility of naturally-aged lead-acid batteries used for safety-related emergency power in nuclear power stations was brought about by battery problems that periodically had been reported in Licensee Event Reports (LERs). The Turkey Point Station had reported cracked and buckled plates in several cells in October 1974 (LER 75-5). The Fitzpatrick Station had reported cracked battery cell cases in October 1977 (LER 77-55) and again in September 1979 (LER 79-59). The Browns Ferry Station had reported a cracked cell leaking a small quantity of electrolyte in July 1981 (LER 81-42). The Indian Point Station had reported cracked and leaking cells in both February (LER 82-7) and April 1982 (LER 82-16); both of these LERs indicated the cracked cells were due to expansion (i.e., growth) of the positive plates

  15. Lessons learned from recent safety related incidents at A Canadian uranium conversion facility

    Jaferi, Jafir

    2013-01-01

    This paper presents the Canadian Nuclear Safety Commission's (CNSC) regulatory requirements for nuclear fuel facility licensees to report any situation or incident that results or is likely to result in a hazard to the health or safety of any person or the environment and to submit its incident investigation report with cause(s) of the incident and corrective actions taken or planned. In addition, the paper presents two recent safety-related incidents that occurred at a uranium conversion facility in Canada along with their consequences, causes, corrective actions and any lessons learned. The first incident resulted in a release of uranium hexafluoride (UF6) inside the UF6 cylinder filling station and the second one resulted in a spill of uranium tetrafluoride (UF 4 ) slurry inside the UF6 plant. Both incidents had no impact on the workers or the environment. (authors)

  16. Fundamentals of a graded approach to safety-related equipment setpoints

    Woodruff, B.A.; Cash, J.S. Jr.; Bockhorst, R.M.

    1993-01-01

    The concept of using a graded approach to reconstitute instrument setpoints associated with safety-related equipment was first presented to the industry by the U.S. Nuclear Regulatory Commission during the 1992 ISA/POWID Symposium in Kansas City, Missouri. The graded approach establishes that the manner in which a utility analyzes and documents setpoints is related to each setpoint's relative importance to safety. This allows a utility to develop separate requirements for setpoints of varying levels of safety significance. A graded approach to setpoints is a viable strategy that minimizes extraneous effort expended in resolving difficult issues that arise when formal setpoint methodology is applied blindly to all setpoints. Close examination of setpoint methodology reveals that the application of a graded approach is fundamentally dependent on the analytical basis of each individual setpoint

  17. Aging related degradation in turbine drives and governors for safety related pumps

    Cox, D.F.

    1991-01-01

    This study is being performed to examine the relationship between time dependent degradation, and current industry practices in the areas of maintenance, surveillance, and operation of stem turbine drive for safety related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized water reactor (PWR) plants, and the Reactor Core Isolation Cooking (RCIC) and High Pressure Coolant Injection (HPCI) systems for Boiling Water Reactor (BWR) facilities. This research has been conducted by examining current information in the Nuclear Plant Reliability Data System (NPRDS), reviewing Licensee Event Reports, thoroughly investigating contacts with operating plant personnel, and by personal observation. This information was reviewed to determine the cause of each reported event and the method of discovery. From this data attempts have been made at determining the predictability of events and possible preventive measures that may be implemented

  18. Smooth handling: the lack of safety-related consumer information in car advertisements.

    Wilson, Nick; Maher, Anthony; Thomson, George; Keall, Michael

    2007-10-01

    To examine the content and trends of safety-related consumer information in magazine vehicle advertisements, as a case study within the worldwide marketing of vehicles. Content analysis of popular current affairs magazines in New Zealand for the 5-year period 2001-2005 was undertaken (n = 514 advertisements), supplemented with vehicle data from official websites. Safety information in advertisements for light passenger vehicles was relatively uncommon with only 27% mentioning one or more of nine key safety features examined (average: 1.7 out of nine features in this 27%). Also included were potentially hazardous features of: speed imagery (in 29% of advertisements), power references (14%), and acceleration data (4%). The speed and power aspects became relatively more common over the 5-year period (p advertisements and vehicle marketing - as already occurs with many other consumer products.

  19. Conference Summary

    Tinkham, M.

    1991-01-01

    This summary will begin with short remarks, trying to recall some of the spirit of the presentations of each of the speakers during the first day, with no attempt at detail or completeness, given the need for a 20:1 compression relative to the original talk. The author hopes these idiosyncratic recollections do not infuriate the speakers too much. Since the speakers on the second day presented such interlocking topics, he simply tries to present some sort of consensus report, to which he adds some comments of his own. The two talks preceding this Summary on the final day dealt with the prospects for applications; since he had no chance to attempt to prepare a proper report on these, he says only a few words about those presentations

  20. Conference summaries

    1987-01-01

    This volume contains summaries of 28 papers presented at the 27. conference of the Canadian Nuclear Association. These papers discuss the general situation of the Canadian nuclear industry and the CANDU reactor; dialogue with the public; the International Atomic Energy Agency; and economic goals and operating lessons. It also contains summaries of 70 papers presented at the 8. conference of the Canadian Nuclear Society, which discuss plant life extension; safety and the environment; reactor physics; thermalhydraulics; risk assessment; the CANDU spacer location and repositioning project; CANDU operations; safety research after Chernobyl; fuel channels; and nuclear technology developments. The individual papers are also available in INIS-mf--13673 (CNA), and INIS-mf--12909 (CNS). (L.L.)

  1. Summary talk

    Harari, H.

    1978-10-01

    A general overview is given in this high energy physics conference summary. Quantum chromodynamics as a theory of strong interactions and studied by experimental tests, SU(2) x U(1) theory of weak and electromagnetic interactions and its experimental tests, weak interactions above 100 GeV, simple unification of weak and electromagnetic interactions, and the grand and the ultimate unifications with extended supergravity are discussed. 28 references

  2. Summary talk

    Johnson, R.C.

    1981-01-01

    In this summary talk some implications of points raised during the Daresbury Study Weekend on heavy-ion reactions are examined and discussed in particular those concerning polarized heavy ions, the connection between analyzing powers and dynamics, transfer reactions, total reaction cross section measurements with polarized beams, and the implications of break-up reaction results for theories of nuclear reactions involving loosely bound projectiles. (U.K.)

  3. Evaluation of the influence of a postulated lubrication oil fire on safety related cables in the top shield platform of PFBR RCB by using FDS Code

    Mangarjuna Rao, P.; Jayasuriya, C.; Nashine, B.K.; Chellapandi, P.; Velusamy, K.

    2010-01-01

    Top deck of Prototype Fast Breeder Reactor (PFBR) primary system houses redundant safety related systems like Control and Safety Rod Drive Mechanisms (CSRDM), Diverse Safety Rod Drive Mechanism (DSRDM), subassembly outlet sodium temperature measurement system and central canal plug. These systems protrude out from the reactor through the Control Plug (CP), which is supported on the Top Shield (TS) of PFBR. Control and instrumentation signal cables and power cables of these safety related systems that are coming out from the CP are routed through Top Shield Platform (TSP, which is concentric with Reactor Vault (RV) at EL 34.1 m above the TS) to the peripheral local instrumentation control centers via the cable junction boxes supported on TS. Influence approach fire hazard analysis (FHA) has been carried out to evaluate the condition of redundant safety related cables under the scenario of a postulated oil fire in the TSP using Fire Dynamics Simulator code (FDS, Version 5). FDS is a computational fluid dynamics (CFD) based fire analysis code and it is developed by National Institute of Standards and Technology (NIST), USA. In this paper the details of the model developed and the results of the analysis carried out are discussed. In TSP, a postulated oil fire scenario with complete inventory of a primary sodium pump (PSP) lubrication oil leak (200 lt) has been considered at 30 m elevation on the TS. Computational model with the geometry of TSP and with other important structural components on TS like PSPs, intermediate heat exchangers (IHXs), large rotating plug (LRP), small rotating plug (SRP), CP and etc. has been developed along with a fire of 1800 kW/m 2 heat release rate in the vicinity of the PSP1. Numerical simulation has been carried out to evaluate this oil fire influence on the typical safety related cables routed at 34 m elevation. It has been found that the surface temperature of the cables that are routed directly above the fire only crosses the ignition

  4. Electric and mechanical basic parameters to elaborate a process for a technical verification of safety related design modifications

    Lamuno Fernandez, Mercedes; La Roca Mallofre, GISEL; Bano Azcon, Alberto

    2010-01-01

    This paper presents a systematic process to check a design in order to achieve all the requirements that regulations demand. Nuclear engineers must verify that a design is done according to the safety requirements, and this paper presents how we have elaborated a process to improve the technical project verification. For a faster, better and easier verification process, here we summarize how to select the electric and mechanical basic parameters, which ensure the correct project verification of safety related design modifications. This process considers different aspects, which guarantee that the design preserves the availability, reliability and functional capability of the Structures, Systems and Components needed to operate the Nuclear Power Station with security. Electric and mechanical reference parameters are identified and discussed as well as others related ones, which are critical to safety. The implementation procedure to develop tasks performed in any company that has a quality plan is a requirement. On the engineering business, it is important not to use the personal criteria to do a technical analysis of a project; although, many times it is the checker's criteria and knowledge responsibility to ensure the correct development of a design modification. Then, the checker capabilities are the basis of the modification verification. This kind of procedure's development is not easy, because in an engineering project with important technical contents, there are multiple scenarios, but lots of them have a common basis. If we can identify the technical common basis of these projects, we will make good project verification but there are many difficulties we can encounter along this process. (authors)

  5. The Fort McMurray Demonstration Project in social marketing: health- and safety-related behaviour among oil sands workers.

    Guidotti, T L; Watson, L; Wheeler, M; Jhangri, G S

    1996-08-01

    This is the first round in a series of surveys conducted in Fort McMurray as part of the Fort McMurray Demonstration Project in social marketing. This component of the survey was intended to focus on the most prominent group of employed workers in the community and to compare their patterns of response with the community as a whole. Respondents to the survey were overwhelmingly male (96%), married (72.9%) and living in households of two to five persons (87.9%). They were predominantly aged 30-44 (55%) and graduates of high school (53.5%). Younger male workers (below age 30) were more likely to have a high school diploma (78.3%) or some additional technical or vocational training (21.7% compared to 12.5% overall) and to be unmarried or separated. Attitudes toward safety-related behaviours were stronger than for respondents from the community as a whole. Approximately 70-100% of all age groups and both sexes showed strong agreement with attitudes involving child car seats and the unacceptability of drinking and driving. These attitudes include strong advocacy of vigorous enforcement of occupational health and safety standards. However, they showed a variability similar to the community as a whole in behaviour at home compared to work, generally reporting more consistent use of personal protection on the job than in their own homes, particularly hearing protection. Even so, they were much less likely to perform stretching and warm-up exercises prior to exertion than community residents in general. The potential may exist to transfer the technology and attitudes from workplace health and safety to community safety. One possible strategy to accomplish this is to involve workers in this industry directly in community initiatives. This strategy may be generalizable to any community in which there are major employers who place a heavy emphasis on risk control and occupational health and safety.

  6. Comparison and Analysis of IEEE 344 and IEC 60980 standards for harmonization of seismic qualification of safety-related equipment

    Lee, Young Ok; Kim, Jong Seog; Seo, Jeong Ho; Kim, Myung Jun

    2011-01-01

    The seismic qualification of safety related equipment in nuclear power plants should demonstrate an equipment's ability to perform its safety function during/or after the time it is subjected to the forces resulting from one SSE. In addition, the equipment must withstand the effects of a number of OBEs, preceding the SSE. IEEE 344 and IEC 60980 present the criteria for establishing procedures demonstrating that the Class 1E equipment can meet its performance requirement during seismic events. Currently, IEEE 344 is used for regulation of nuclear power plant in the United State whereas IEC 60980 is mainly used in Europe. In particular, NPPs of France and China apply with RCC-E and GB that are domestic standards, respectively. Equipment supplier and Utility have difficulties because of different applicable standards. Equipment supplier to export S/R components/equipment to other standard area performs additional seismic qualification. For example, equipment are qualifies according to IEC 60980, RCC-E, GB although they have been qualified in accordance with IEEE 344. Also, utility to attempt power up-rate, life extension of NPP constructed under rules of RCC-E such as Ulchin NPP 1 and 2 has similar difficulties. RCC-E endorses IEC 60980 and GB is almost same as IEC 60980 except minor difference of earthquake environment definition. Therefore this paper surveys the similarities and differences between IEEE 344 and IEC 60980. In addition, this paper considers how the two sets of standards may be used in a complementary fashion to be possible using one or the other standard area

  7. Executive summary

    2002-01-01

    On 18 May 2001, the Finnish Parliament ratified the Decision in Principle on the final disposal facility for spent nuclear fuel at Olkiluoto, within the municipality of Eurajoki. The Municipality Council and the government has made positive decisions earlier, at the end of 2000, and in compliance with the Nuclear Energy Act, Parliament's ratification was then required. The decision is valid for the spent fuel generated by the existing Finnish nuclear power plants and means that the construction of the final disposal facility is considered to be in line with the overall good of society. Earlier steps included, amongst others, the approval of the technical project by the Safety Authority. Future steps include construction of an underground rock characterisation facility, ONKALO (2003-2004), and application for separate construction and operating licences for the final disposal facility (from about 2010). How did this political and societal decision come about? The FSC Workshop provided the opportunity to present the history leading up to the Decision in Principle (DiP), and to examine future perspectives with an emphasis on stakeholder involvement. This Executive Summary gives an overview of the presentations and discussions that took place at the workshop. It presents, for the most part, a factual account of the individual presentations and of the discussions that took place. It relies importantly on the notes that were taken at the meeting. Most materials are elaborated upon in a fuller way in the texts that the various speakers and session moderators contributed for these proceedings. The structure of the Executive Summary follows the structure of the workshop itself. Complementary to this Summary and also provided with this document, is a NEA Secretariat's perspective aiming to place the results of all discussions, feedback and site visit into an international perspective. (authors)

  8. Executive summary

    1981-02-01

    This paper is an 'executive summary' of work undertaken to review proposals for transport, handling and emplacement of high level radioactive wastes in an underground repository, appropriate to the U.K. context, with particular reference to: waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transportation within the repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; and costs. The paper contains a section on general conclusions and recommendations. (U.K.)

  9. Summary guidelines

    Halsnaes, K.; Painuly, J.P.; Turkson, J.; Meyer, H.J.; Markandya, A.

    1999-09-01

    This document is a summary version of the methodological guidelines for climate change mitigation assessment developed as part of the Global Environment Facility (GEF) project Economics of Greenhouse Gas Limitations; Methodological Guidelines. The objectives of this project have been to develop a methodology, an implementing framework and a reporting system which countries can use in the construction of national climate change mitigation policies and in meeting their future reporting obligations under the FCCC. The methodological framework developed in the Methodological Guidelines covers key economic concepts, scenario building, modelling tools and common assumptions. It was used by several country studies included in the project. (au) 13 refs.

  10. Conference summaries

    1988-01-01

    This volume contains conference summaries of the 28. annual conference of the Canadian Nuclear Association, and the 9. annual conference of the Canadian Nuclear Society. Topics of discussion include: power reactors; fuel cycles; nuclear power and public understanding; future trends; applications of nuclear technology; CANDU reactors; operational enhancements; design of small reactors; accident behaviour in fuel channels; fuel storage and waste management; reactor commissioning/decommissioning; nuclear safety experiments and modelling; the next generation reactors; advances in nuclear engineering education in Canada; safety of small reactors; current position and improvements of fuel channels; current issues in nuclear safety; and radiation applications - medical and industrial

  11. Replacement cross-site transfer system project W-058 safety class upgrade summary report

    Schlosser, R.L.

    1998-01-01

    This report evaluates the design of the replacement cross-site transfer system structures, systems, and components for safety related applications as defined in the Tank Waste Remediation Systems Basis for Interim Operations

  12. Testing existing software for safety-related applications. Revision 7.1

    Scott, J.A.; Lawrence, J.D.

    1995-12-01

    The increasing use of commercial off-the-shelf (COTS) software products in digital safety-critical applications is raising concerns about the safety, reliability, and quality of these products. One of the factors involved in addressing these concerns is product testing. A tester's knowledge of the software product will vary, depending on the information available from the product vendor. In some cases, complete source listings, program structures, and other information from the software development may be available. In other cases, only the complete hardware/software package may exist, with the tester having no knowledge of the internal structure of the software. The type of testing that can be used will depend on the information available to the tester. This report describes six different types of testing, which differ in the information used to create the tests, the results that may be obtained, and the limitations of the test types. An Annex contains background information on types of faults encountered in testing, and a Glossary of pertinent terms is also included. This study is pertinent for safety-related software at reactors

  13. Analysis of Paks NPP Personnel Activity during Safety Related Event Sequences

    Bareith, A.; Hollo, Elod; Karsa, Z.; Nagy, S.

    1998-01-01

    Within the AGNES Project (Advanced Generic and New Evaluation of Safety) the Level-1 PSA model of the Paks NPP Unit 3 was developed in form of a detailed event tree/fault tree structure (53 initiating events, 580 event sequences, 6300 basic events are involved). This model gives a good basis for quantitative evaluation of potential consequences of actually occurred safety-related events, i.e. for precursor event studies. To make these studies possible and efficient, the current qualitative event analysis practice should be reviewed and a new additional quantitative analysis procedure and system should be developed and applied. The present paper gives an overview of the method outlined for both qualitative and quantitative analyses of the operator crew activity during off-normal situations. First, the operator performance experienced during past operational events is discussed. Sources of raw information, the qualitative evaluation process, the follow-up actions, as well as the documentation requirements are described. Second, the general concept of the proposed precursor event analysis is described. Types of modeled interactions and the considered performance influences are presented. The quantification of the potential consequences of the identified precursor events is based on the task-oriented, Level-1 PSA model of the plant unit. A precursor analysis system covering the evaluation of operator activities is now under development. Preliminary results gained during a case study evaluation of a past historical event are presented. (authors)

  14. Structure soil structure interaction effects: Seismic analysis of safety related collocated concrete structures

    Joshi, J.R.

    2000-01-01

    The Process, Purification and Stack Buildings are collocated safety related concrete shear wall structures with plan dimensions in excess of 100 feet. An important aspect of their seismic analysis was the determination of structure soil structure interaction (SSSI) effects, if any. The SSSI analysis of the Process Building, with one other building at a time, was performed with the SASSI computer code for up to 50 frequencies. Each combined model had about 1500 interaction nodes. Results of the SSSI analysis were compared with those from soil structure interaction (SSI) analysis of the individual buildings, done with ABAQUS and SASSI codes, for three parameters: peak accelerations, seismic forces and the in-structure floor response spectra (FRS). The results may be of wider interest due to the model size and the potential applicability to other deep soil layered sites. Results obtained from the ABAQUS analysis were consistently higher, as expected, than those from the SSI and SSSI analyses using the SASSI. The SSSI effect between the Process and Purification Buildings was not significant. The Process and Stack Building results demonstrated that under certain conditions a massive structure can have an observable effect on the seismic response of a smaller and less stiff structure

  15. Identification of potential safety-related incidents applicable to a breeder fuel reprocessing plant

    Perkins, W.C.

    1980-01-01

    The current emphasis on safety in all phases of the nuclear fuel cycle requires that safety features be identified and included in designs of nuclear facilities at the earliest possible stage. A popular method for the early identification of these safety features is the Preliminary Hazards Analysis. An extension of this analysis is to illustrate the nature of a hazard by its effects in accident situations, that is, to identify what are called safety-related incidents. Some useful tools are described which have been used at the Savannah River Laboratory, SRL, to make Preliminary Hazards Analyses as well as safety analyses of facilities for processing spent nuclear fuels from both power and production reactors. These tools have also been used in safety studies of waste handling operations at the Savannah River Plant. The tools are the SRL Incidents Data Bank and the What If meeting. The application of this methodology to a proposed facility which has breeder fuel reprocessing capability, the Hot Experimental Facility (HEF) is illustrated

  16. Priority ranking of safety-related systems for structural assessment at Savannah River Site

    Kao, G.C.; Daugherty, W.L.; Barnes, D.M.

    1993-01-01

    In order to extend the service life of safety related structures and systems in a logical manner, a Structural Enhancement Program was initiated to evaluate the structural integrity of eight systems, namely: cooling water system, emergency cooling system, moderator recovery system, supplementary safety system, water removal system, service raw water system, service clarified water system, and river water system. Since the level of importance of each system to reactor operations varies from one system to another, the scope of structural integrity evaluation for each system should be prioritized accordingly. This paper presents the assessment of system priority for structural evaluation based on a ranking methodology and specifies the level of structural evaluation consistent with the established priority. The effort was undertaken by a five-member panel representing four major disciplines, including: structures, reactor engineering/operations, risk management, and materials. The above systems were divided into a total of thirty-five subsystems. These subsystems were then ranked with six attributes, namely: safety classification, degradation mechanisms, difficulty of replacement, failure mode, radiation dose to workers, and consequence of failure. Each attribute was assigned a set of consequences or events with corresponding weighting scores. The results of the ranking process yielded two groups of subsystems, categorized as Priority I and II subsystems. The level of structural assessment was then formulated accordingly. The prioritized approach will allow more efficient allocation of resources, so that the Structural Enhancement Program can be implemented in a cost-effective and efficient manner

  17. Priority ranking of safety-related systems for structural enhancement assessment at Savannah River Site

    Kao, G.C.; Daugherty, W.L.; Barnes, D.M.

    1992-09-01

    In order to extend the service life of safety related structures and systems in a logical manner, a Structural Enhancement Program was initiated to evaluate the structural integrity of eight (8) systems, namely: Cooling Water System, Emergency Cooling System, Moderator Recovery System supplementary Safety System, Water Removal System, Service Raw Water System, Service Clarified Water System, and River Water System. Since the level of importance of each system to reactor operations varies from one system to another, the scope of structural integrity evaluation for each system should be prioritized accordingly. This paper presents the assessment of system priority for structural evaluation based on a ranking methodology and specifies the level of structural evaluation consistent with the established priority. The effort was undertaken by a five-member panel representing four (4) major disciplines, including. structures, reactor engineering/operations, risk management and materials. The above systems were divided into a total of thirty-five (35) subsystem. These subsystems were then ranked with six (6) attributes, namely: Safety Classification, Degradation Mechanisms, Difficulty of Replacement, Failure Mode, Radiation Dose to Workers and Consequence of Failure. Each attribute was assigned a set of consequences or events with corresponding weighting scores. The results of the ranking process yielded two groups of subsystems, categorized as Priority I and II subsystems. The level of structural assessment was then formulated accordingly. The prioritized approach will allow more efficient allocation of resources, so that the Structural Enhancement Program can be implemented in a cost-effective and efficient manner

  18. Assessment of modular construction for safety-related structures at advanced nuclear power plants

    Braverman, J.; Morante, R.; Hofmayer, C.

    1997-03-01

    Modular construction techniques have been successfully used in a number of industries, both domestically and internationally. Recently, the use of structural modules has been proposed for advanced nuclear power plants. The objective in utilizing modular construction is to reduce the construction schedule, reduce construction costs, and improve the quality of construction. This report documents the results of a program which evaluated the proposed use of modular construction for safety-related structures in advanced nuclear power plant designs. The program included review of current modular construction technology, development of licensing review criteria for modular construction, and initial validation of currently available analytical techniques applied to concrete-filled steel structural modules. The program was conducted in three phases. The objective of the first phase was to identify the technical issues and the need for further study in order to support NRC licensing review activities. The two key findings were the need for supplementary review criteria to augment the Standard Review Plan and the need for verified design/analysis methodology for unique types of modules, such as the concrete-filled steel module. In the second phase of this program, Modular Construction Review Criteria were developed to provide guidance for licensing reviews. In the third phase, an analysis effort was conducted to determine if currently available finite element analysis techniques can be used to predict the response of concrete-filled steel modules

  19. Development of a Method for Quantifying the Reliability of Nuclear Safety-Related Software

    Yi Zhang; Golay, Michael W.

    2003-01-01

    The work of our project is intended to help introducing digital technologies into nuclear power into nuclear power plant safety related software applications. In our project we utilize a combination of modern software engineering methods: design process discipline and feedback, formal methods, automated computer aided software engineering tools, automatic code generation, and extensive feasible structure flow path testing to improve software quality. The tactics include ensuring that the software structure is kept simple, permitting routine testing during design development, permitting extensive finished product testing in the input data space of most likely service and using test-based Bayesian updating to estimate the probability that a random software input will encounter an error upon execution. From the results obtained the software reliability can be both improved and its value estimated. Hopefully our success in the project's work can aid the transition of the nuclear enterprise into the modern information world. In our work, we have been using the proprietary sample software, the digital Signal Validation Algorithm (SVA), provided by Westinghouse. Also our work is being done with their collaboration. The SVA software is used for selecting the plant instrumentation signal set which is to be used as the input the digital Plant Protection System (PPS). This is the system that automatically decides whether to trip the reactor. In our work, we are using -001 computer assisted software engineering (CASE) tool of Hamilton Technologies Inc. This tool is capable of stating the syntactic structure of a program reflecting its state requirements, logical functions and data structure

  20. Seismic analysis for safety related structures of 900MWe PWR NPP

    Liu Wei

    2002-01-01

    Nuclear Power Plant aseismic design becomes more and more important in China due to the fact that China is a country where earthquakes occur frequently and most of plants arc unavoidably located in seismic regions. Therefore, Chinese nuclear safety authority and organizations have worked out a series of regulations and codes related to NPP anti-seismic design taking account of local conditions. The author presents here an example of structural anti-seismic design of 90GM We PWR NPP which is comprised of: ground motion input, including the principles for ground motion determination and time history generation; soil and upper-structure modelling, presenting modeling procedures and typical models of safety related buildings such as Reactor Building, Nuclear Auxiliary Building and Fuel Building; soil-structure interaction analysis; and in-structure response analysis and floor response spectrum generation. With this example, the author intends to give an overview of Chinese practice in NPP structure anti-seismic design such as the main procedures to be followed and the codes and regulations to be respected. (author)

  1. Achieving Minimum Clinically Important Difference in Oxford Knee Score and Short Form-36 Physical Component Summary Is Less Likely with Single-Radius Compared with Multiradius Total Knee Arthroplasty in Asians.

    Lee, Wu Chean; Bin Abd Razak, Hamid Rahmatullah; Allen, John Carson; Chong, Hwei Chi; Tan, Hwee Chye Andrew

    2018-04-10

    Single-radius (SR) and multiradius (MR) total knee arthroplasties (TKAs) have produced similar outcomes, albeit most studies originate from Western nations. There are known knee kinematic differences between Western and Asian patients after TKA. The aim of this study is to compare the short-term patient-reported outcome measures (PROMs) of SR-TKA versus MR-TKA in Asians. Registry data of 133 SR-TKA versus 363 MR-TKA by a single surgeon were analyzed. Preoperative and 2-year postoperative range of motion (ROM) and PROMs were compared with Student's t -test and Mann-Whitney U-test. Logistic regression model was used to evaluate the odds of SR-TKA or MR-TKA achieving the minimum clinically important difference (MCID) of studied outcomes. Patients in both groups had similar age (65.7 ± 7.6 vs. 65.8 ± 8.2 years; p  = 0.317), gender proportion (71% females vs. 79% females; p  = 0.119), and ethnic distribution (80% Chinese vs. 84% Chinese; p  = 0.258). Preoperatively, there were no statistically significant differences between both groups for ROM, Knee Society Score (KSS), Oxford Knee Score (OKS), and Short Form (SF)-36 scores. At 2 years, all outcomes were statistically similar or failed to achieve a difference of MCID. Controlling for all preoperative variables, SR-TKA has significantly lower odds of achieving MCID for OKS (odds ratio [OR]: 0.275, 95% confidence interval [CI]: 0.114-0.663; p  = 0.004) and SF-36 Physical Component Summary (PCS) (OR: 0.547; 95% CI: 0.316-0.946; p  = 0.031) compared with MR-TKA. In conclusion, there are no significant differences in the absolute PROMs between SR-TKA and MR-TKA at 2 years following TKA in Asians. However, SR-TKA has significantly lower odds of achieving the MCID for OKS and SF-36 PCS. Thieme Medical Publishers 333 Seventh Avenue, New York, NY 10001, USA.

  2. Evaluating North Carolina Food Pantry Food Safety-Related Operating Procedures.

    Chaifetz, Ashley; Chapman, Benjamin

    2015-11-01

    Almost one in seven American households were food insecure in 2012, experiencing difficulty in providing enough food for all family members due to a lack of resources. Food pantries assist a food-insecure population through emergency food provision, but there is a paucity of information on the food safety-related operating procedures used in the pantries. Food pantries operate in a variable regulatory landscape; in some jurisdictions, they are treated equivalent to restaurants, while in others, they operate outside of inspection regimes. By using a mixed methods approach to catalog the standard operating procedures related to food in 105 food pantries from 12 North Carolina counties, we evaluated their potential impact on food safety. Data collected through interviews with pantry managers were supplemented with observed food safety practices scored against a modified version of the North Carolina Food Establishment Inspection Report. Pantries partnered with organized food bank networks were compared with those that operated independently. In this exploratory research, additional comparisons were examined for pantries in metropolitan areas versus nonmetropolitan areas and pantries with managers who had received food safety training versus managers who had not. The results provide a snapshot of how North Carolina food pantries operate and document risk mitigation strategies for foodborne illness for the vulnerable populations they serve. Data analysis reveals gaps in food safety knowledge and practice, indicating that pantries would benefit from more effective food safety training, especially focusing on formalizing risk management strategies. In addition, new tools, procedures, or policy interventions might improve information actualization by food pantry personnel.

  3. Development of a Method for Quantifying the Reliability of Nuclear Safety-Related Software

    Yi Zhang; Michael W. Golay

    2003-10-01

    The work of our project is intended to help introducing digital technologies into nuclear power into nuclear power plant safety related software applications. In our project we utilize a combination of modern software engineering methods: design process discipline and feedback, formal methods, automated computer aided software engineering tools, automatic code generation, and extensive feasible structure flow path testing to improve software quality. The tactics include ensuring that the software structure is kept simple, permitting routine testing during design development, permitting extensive finished product testing in the input data space of most likely service and using test-based Bayesian updating to estimate the probability that a random software input will encounter an error upon execution. From the results obtained the software reliability can be both improved and its value estimated. Hopefully our success in the project's work can aid the transition of the nuclear enterprise into the modern information world. In our work, we have been using the proprietary sample software, the digital Signal Validation Algorithm (SVA), provided by Westinghouse. Also our work is being done with their collaboration. The SVA software is used for selecting the plant instrumentation signal set which is to be used as the input the digital Plant Protection System (PPS). This is the system that automatically decides whether to trip the reactor. In our work, we are using -001 computer assisted software engineering (CASE) tool of Hamilton Technologies Inc. This tool is capable of stating the syntactic structure of a program reflecting its state requirements, logical functions and data structure.

  4. Conference summaries

    Reynolds, Tim [Inta Communication Limited for European Service Network/ DG Research, Trillium House, 32 New Street, St. Neots, Cambridge PE19 1AJ (United Kingdom)

    2004-07-01

    The summaries were derived from presentations, interviews and discussions at the conference. The summaries are given at two levels, overall for the conference and for specific sessions as follows: 1) Overall Conference: 'A Sound Scientific Basis for Serious Decisions; 2) Sessions on EC Policy and Socio-Political Issues: 'Promoting Safety and Protecting Society'; 3) Session on P and T: 'Partitioning and Transmutation: A Technical Fix or Technical Training?'; 4) Sessions on Geological Disposal and Research Networking: 'No Technical Barriers to Geological Disposal'. First an overall summary of Euradwaste '04 is presented. Significant progress was made on the technical and scientific basis for geological disposal of radioactive waste during the European Commission's Fifth EURATOM Framework Programme for Research (FP5). Deep geological disposal is technically feasible now and can demonstrate the guarantees of long-term isolation and protection of the public. In parallel, socio-political studies have produced methodologies for constructive dialogue with potential host communities that reflect the honesty and openness expected by a democratic society. A harmonized legislative framework for nuclear safety and waste disposal across the enlarged European Union is currently being discussed. Disposal in deep (> 300 metre) geological repositories, the favoured strategy in Europe for long-lived high-level radioactive waste, is now possible. The Sessions on EC Policy and Socio-Political Issues are summarized as follows. The opening day of Euradwaste '04 focused on European Commission policy, including the proposed Directives on disposal of radioactive waste and nuclear safety and socio-political aspects including governance and decision making, public perception/acceptance of waste disposal and its sustainability. A decision on the proposed package will now be made after Union enlargement. Public agreement on the siting of

  5. PGDP [Paducah Gaseous Diffusion Plant]-UF6 handling, sampling, analysis and associated QC/QA and safety related procedures

    Harris, R.L.

    1987-01-01

    This document is a compilation of Paducah Gaseous Diffusion Plant procedures on UF 6 handling, sampling, and analysis, along with associated QC/QA and safety related procedures. It was assembled for transmission by the US Department of Energy to the Korean Advanced Energy Institute as a part of the US-Korea technical exchange program

  6. Executive Summary

    2014-01-01

    Each session of the workshop consisted of a number of presentations followed by a panel discussion moderated by the session Chairs. A summary of each session and subsequent discussion that ensued are provided. Session 1: National approaches for long term interim storage facilities. Seven papers were presented during this session by representatives of research institutes in USA (EPRI) and in Norway (IFE), governmental authorities for the nuclear industry in Finland (STUK) and Slovak Republic (UJD), technical support organizations in Germany (GRS) and France (IRSN) and the public company in charge of waste management in Spain (ENRESA). The papers discussed the national policy, the regulatory framework and the current situation for storage of SF and HLW in various European countries (Germany, Spain, Finland, Norway and Slovak Republic). The main activities the EPRI is undertaking to establish the technical bases for extended (long-term) storage and the IRSN's definition of the safety principles and objectives for new storage facilities regarding long-term storage are also discussed. Session 2: Safety requirements, regulatory framework and implementation issues. Eleven papers were presented during this session by representatives from international groups (the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency (NEA)), representatives of regulatory bodies from the United States (Nuclear Regulatory Commission) and Germany (Bundesamt fuer Strahlenschutz (BfS)), German Federal Institute for Materials Research and Testing (BAM), operators from the United Kingdom (Sellafield Limited, UK) and France (EDF), vendors (AREVA), and representatives of TSOs in Germany (TUV and Oko-Institut). Session 3: Technical issues and operational experience, needs for R and D. This session was chaired by Karl Wasinger (AREVA, Germany) and Fumihisa Nagase (JAEA, Japan). Ten papers were presented during this session by representatives of research institutes of Japan (CRIEPI

  7. Session summaries

    Sudo, Y.

    2002-01-01

    In the summary session, possible international activities in the field of basic studies on high-temperature engineering were discussed within the framework of the OECD/NEA Nuclear Science Committee (NSC). It was recommended to include topics relevant to fission-product behaviour and safety issues of HTGR in next meeting, in addition to the topics discussed in this meeting. The chairperson of the last session summarised the recommendations to be presented to the NSC into the following five topics as possible international activities: - Basic studies on behaviour of irradiated graphite/carbon and ceramic materials including their composites under both operation and storage conditions. - Development of in-core material characterisation and instrumentation methods. - Improvement in material properties through high-temperature irradiation. - Basic studies on HTGR fuel fabrication and performance including fission-product release. - Basic studies on safety issues of HTGR. It was also recommended that a further information exchange meeting focused on the organisation of the interactive collaboration activity with regard to the above topics be planned in 2003, tentatively in Oarai, Japan. (author)

  8. Management summary

    Anon.

    1977-01-01

    A most pressing problem in many environmental assessments is determining the extent of contamination to the biosphere resulting from a given activity. This could result from the planned or accidental release of a contaminant to the environment and its subsequent transport through air, water, or the food chain. In either case, three critical questions need to be raised in the environmental assessment of the problem: Where are the contaminants; When will they arrive at a specific location; How much of the contaminant will be at the point of uptake. The location of the contaminant is important, since a contaminant isolated from man both now and in the future may represent little hazard, even in rather large quantities. Under other conditions, small amounts of contaminants arriving at critical locations over a short period may involve severe hazard. The problem of location is simplified by concentrating on those places where the contaminants will interface with the biosphere. Applications in evaluating the consequences of ground water contamination are discussed. Environmental consequences or impacts are most effectively and efficiently communicated by: Blending extensive technical results and reducing them to simple summary relationships, i.e., the arrival distributions; focusing on the arrival distributions as the central theme of communication efforts; and determining quantitative consequences or impacts to the biosphere through use of the arrival distributions. Appropriately applied, these principles can reduce a voluminous impact statement on subsurface pollution to a few pages that are directly useful to decision-makers and the public

  9. Theory summary

    Tang, W.M.

    2001-01-01

    This is a summary of the advances in magnetic fusion energy theory research presented at the 17th International Atomic Energy Agency Fusion Energy Conference from 19 24 October, 1998 in Yokohama, Japan. Theory and simulation results from this conference provided encouraging evidence of significant progress in understanding the physics of thermonuclear plasmas. Indeed, the grand challenge for this field is to acquire the basic understanding that can readily enable the innovations which would make fusion energy practical. In this sense, research in fusion energy is increasingly able to be categorized as fitting well the 'Pasteur's Quadrant' paradigm, where the research strongly couples basic science ('Bohr's Quadrant') to technological impact ('Edison's Quadrant'). As supported by some of the work presented at this conference, this trend will be further enhanced by advanced simulations. Eventually, realistic three-dimensional modeling capabilities, when properly combined with rapid and complete data interpretation of results from both experiments and simulations, can contribute to a greatly enhanced cycle of understanding and innovation. Plasma science theory and simulation have provided reliable foundations for this improved modeling capability, and the exciting advances in high-performance computational resources have further accelerated progress. There were 68 papers presented at this conference in the area of magnetic fusion energy theory

  10. Substance use and social, health and safety-related factors among fatally injured drivers.

    Karjalainen, Karoliina; Blencowe, Tom; Lillsunde, Pirjo

    2012-03-01

    The aim of this study was to examine different socio-demographic, health and safety-related factors, and psychoactive substance use among fatally injured drivers in road traffic accidents in Finland during 2006-2008. An accident information register maintained by the Traffic Safety Committee of Insurance Companies (VALT) of the Finnish Motor Insurers' Centre was used as basic data, and the basic data were complemented with further toxicological analytical information retrieved from autopsy reports from the Department of Forensic Medicine, Helsinki University. The data included all the drivers (n=556) who were driving a motor vehicle and who died in a road traffic accident in Finland during 2006-2008. Of all the 556 fatally injured drivers 43% (n=238) had psychoactive substance findings. 51% (n=121) of substance positive drivers had a finding for alcohol only, the rest had a finding for one or more illicit/medicinal drugs impairing driving ability, and possibly also alcohol. Fatally injured drivers with alcohol findings were significantly younger (mean age 34 years) than sober drivers (mean age 44 years) or drivers with findings for drugs (mean age 45 years). Socio-demographic background did not differ substantially among drunken/drugged and sober drivers, although drivers with alcohol findings had a slightly lower education and socioeconomic position. Previous substance abuse problems were highly prevalent among drivers with substance findings and mental or both mental and physical health problems were more common among drivers with drug findings. The non-use of safety equipment and driving at a high speed were more common among fatally injured drivers with substance findings. Substance abuse and mental health problems, as well as reckless driving behavior were more pronounced among fatally injured drivers with substance findings when compared to sober drivers. Thus, prevention and early intervention concerning substance abuse, mental health problems and DUI are

  11. Safety related design and economic aspects of HTGRs. Proceedings of a technical committee meeting

    NONE

    2001-04-01

    The purpose of the Technical Committee Meeting (TCM) was to provide the opportunity to review the status of design and development activities associated with safety related and economic aspects of HTGRs, and to identify pathways which may provide the opportunity for international cooperation in addressing these issues. The HTGR, as a nuclear heat source for the safe, economic and efficient production of electricity and high temperature industrial processes has, within the past few years, become a significantly increasing influence in the future of nuclear power. Nuclear test facilities with the capability of achieving core outlet temperatures to 950 deg. C are presently under construction in China and Japan. These plants will be utilized to support HTGR research and development activities, including electricity generation via the gas turbine and validation of high temperature process heat applications. Also, major development programmes focusing on the generation of electricity through the direct cycle gas turbine are in progress by ESKOM, the state electric utility of South Africa, and by a consortium of organizations from the Russian Federation, USA, France and Japan. Other national programmes focusing on research and development of the HTGR are underway including the Netherlands, where an evaluation is being completed on a heat and power co-generation plant utilizing a small direct cycle HTR; in Germany, where the primary focus is centered on basic issues of reactor safety and innovative reactor technology; in Indonesia with the evaluation of process heat applications such as coal liquefaction, hydrogen production and high temperature reforming of methane; and in the USA with the recent re-introduction of national support for the HTGR specifically directed to the burning of weapons plutonium. The status information presented in several of the papers is as of the time of drafting. Thus other later material should be referenced for more current status information

  12. Safety related design and economic aspects of HTGRs. Proceedings of a technical committee meeting

    2001-04-01

    The purpose of the Technical Committee Meeting (TCM) was to provide the opportunity to review the status of design and development activities associated with safety related and economic aspects of HTGRs, and to identify pathways which may provide the opportunity for international cooperation in addressing these issues. The HTGR, as a nuclear heat source for the safe, economic and efficient production of electricity and high temperature industrial processes has, within the past few years, become a significantly increasing influence in the future of nuclear power. Nuclear test facilities with the capability of achieving core outlet temperatures to 950 deg. C are presently under construction in China and Japan. These plants will be utilized to support HTGR research and development activities, including electricity generation via the gas turbine and validation of high temperature process heat applications. Also, major development programmes focusing on the generation of electricity through the direct cycle gas turbine are in progress by ESKOM, the state electric utility of South Africa, and by a consortium of organizations from the Russian Federation, USA, France and Japan. Other national programmes focusing on research and development of the HTGR are underway including the Netherlands, where an evaluation is being completed on a heat and power co-generation plant utilizing a small direct cycle HTR; in Germany, where the primary focus is centered on basic issues of reactor safety and innovative reactor technology; in Indonesia with the evaluation of process heat applications such as coal liquefaction, hydrogen production and high temperature reforming of methane; and in the USA with the recent re-introduction of national support for the HTGR specifically directed to the burning of weapons plutonium. The status information presented in several of the papers is as of the time of drafting. Thus other later material should be referenced for more current status information

  13. Safety related maintenance in the framework of the reliability centered maintenance concept

    1992-07-01

    Elevated safety requirements and ever increasing costs of maintenance of nuclear power plants stimulate the interest in different methods and approaches to optimize maintenance activities. Among different concepts, the Reliability Centered Maintenance (RCM) as an approach to improve Preventive Maintenance (PM) programmes is being widely discussed an applied in several IAEA Member States. In order to summarize basic principles and current implementation of the RCM, the IAEA organized a Consultants Meeting in November 1990. The report prepared during that meeting was discussed during the Technical Committee Meeting (TCM) held in May 1991. Numerous technical presentations as well as panel and plenary discussions took place at the TCM. This document contains the report of the Consultants Meeting (modified to include comments of the TCM), a summary of the most important discussions as well as all 14 papers presented at the TCM

  14. Executive summary

    2002-01-01

    systems. The scope of the workshop comprised reactor physics, fuel performance and fuel material technology, thermal-hydraulics, core behaviour and fuel cycle of advanced reactors with different types of fuels or fuel lattices. Reactor types considered were water-cooled, high-temperature gas-cooled and fast spectrum reactors as well as hybrid reactors with fast and thermal neutron spectra. The emphasis was on innovative concepts and issues related to the reactor and fuel. The workshop concluded with a wide-ranging panel discussion which considered some difficult questions from which it is hoped that some recommendations for future priorities can be derived. A record of the discussion is included at the end of this summary. (author)

  15. Investigation of potential fire-related damage to safety-related equipment in nuclear power plants

    Wanless, J.

    1985-11-01

    Based on a review of vendor information, fire damage reports, equipment qualification and hydrogen burn test results, and material properties, thirty-three types of equipment found in nuclear power plants were ranked in terms of their potential sensitivity to fire environments. The ranking considered both the functional requirements and damage proneness of each component. A further review of the seven top-ranked components was performed, considering the relative prevalence and potential safety significance of each. From this, relays and hand switches dominate as first choices for fire damage testing with logic equipment, power supplies, transmitters, and motor control centers as future candidates

  16. IE Information Notice No. 85-08: Industry experience on certain materials used in safety-related equipment

    Jordan, E.L.

    1992-01-01

    This information notice is being issued to provide licensees and construction permit holders with information pertaining to the behavior of certain materials used in safety-related equipment. The materials, as described below, were observed to have the potential of degrading the operability of safety-related equipment. These observations were made during environmental qualification testing and/or during routine inspection of in-service equipment. The notice describes the following: elastomeric seals used in personnel air locks for the reactor containment systems; epoxy phenolic coating applied to the lower portion of the interior surface of diesel oil storage tanks; the use of Viton elastomer as the seal material in hydrogen recombiner applications; and environmental qualification of ASCO NP valves with Viton and ethylene propylene parts

  17. A Study on Performance Evaluation of Safety-Related Protective Coating for Yonggwang Unit 1 and 2

    Kim, Sung Young; Kim, Young Bum; Lee, Won Sang

    2010-01-01

    Protective coating inside nuclear power plants could find its origin from NRC Reg. Guide 1.82(Rev. 3) regarding current issue for the regulation of foreign materials inside containment building. The current issue for the regulation of foreign materials inside containment considered/determined the current issues only regarding the blockage of sump screen by foreign materials such as coating material, insulator, and other materials, while safety-related coating is separately managed by NRC Reg. Guide 1.54(Rev. 1). In this study, we performed field walk-down to evaluate the as-is condition of protective coating inside containment building which was classified as for structure and for equipment with applying the requirement for safety-related coating

  18. Sophisticated Calculation of the 1oo4-architecture for Safety-related Systems Conforming to IEC61508

    Hayek, A; Al Bokhaiti, M; Schwarz, M H; Boercsoek, J

    2012-01-01

    With the publication and enforcement of the standard IEC 61508 of safety related systems, recent system architectures have been presented and evaluated. Among a number of techniques and measures to the evaluation of safety integrity level (SIL) for safety-related systems, several measures such as reliability block diagrams and Markov models are used to analyze the probability of failure on demand (PFD) and mean time to failure (MTTF) which conform to IEC 61508. The current paper deals with the quantitative analysis of the novel 1oo4-architecture (one out of four) presented in recent work. Therefore sophisticated calculations for the required parameters are introduced. The provided 1oo4-architecture represents an advanced safety architecture based on on-chip redundancy, which is 3-failure safe. This means that at least one of the four channels have to work correctly in order to trigger the safety function.

  19. Generic requirements specification for qualifying a commercially available PLC for safety-related applications in nuclear power plants. Final report

    Ostenso, A.; May, R.

    1996-12-01

    This is a specification for qualifying a commercially available PLC for application to safety systems in nuclear power plants. The specifications are suitable for evaluating a particular PLC product line as a platform for safety-related applications, establishing a suitable qualification test program, and confirming that the manufacturer has a quality assurance program that is adequate for safety-related applications or is sufficiently complete that, with a reasonable set of compensatory actions, it can be brought into conformance. The specification includes requirements for: (1) quality assurance measures applied to the qualification activities, (2) documentation to support the qualification, and (3) documentation to provide the information needed for applying the qualified PLC platform to a specific application. The specifications are designed to encompass a broad range of safety applications; however, qualifying a particular platform for a different range of applications can be accomplished by appropriate adjustments to the requirements

  20. Development of integrated D/B system for the safety-related structures in nuclear power plant

    Cho, M. S.; Song, Y. C.; Lee, J. S.; Choi, W. S.

    2002-01-01

    The integrated D/B system is developed for digitalizing the history of the safety-related structures of nuclear power plant. It have 5 database which are consist of Generals, Structural and Design, Materials, Construction, Aging and repair information D/B. For efficient operation of the system, we are to set up the outline of the system, find out data field for target structures, and develop utilities. Utilities will be the aging and repair data management program, the close examination management program, the data search engine with various options which help users to find the information quickly, and the data management program restoring, updating and exchanging input data. Development of the integrated D/B system of the safety-related structures will contribute to management of the structures of nuclear power plant with advanced technology

  1. Executive Summary

    2013-01-01

    countries (MC) and concerned both initial safety assessment of new facilities and reassessment of existing ones (periodic safety review). It also considered trends of future improvement of safety assessment techniques. The workshop was organised in an opening session, four technical sessions, one special session and a conclusion session. The technical sessions were focussed on: - General approach including human aspects (9 papers); - Front end facilities (5 papers); - Chemical hazards - release limits (6 papers); and - Back end facilities (6 papers). In addition, a special session (4 presentations) was held to discuss the lessons learnt for FCFs from the Fukushima accident in Japan. The workshop ended with an organized site visit to Cameco Corporation's Port Hope Conversion Facility in Port Hope, Ontario on the last day of the workshop. This paper presents the Summary of the technical and special sessions, the General Conclusions and Recommendation of the workshop and some future directions

  2. Summary Record

    2008-01-01

    (II-3) - Uncertainty analysis of the steady state benchmark. It should be recognized that the purpose of this benchmark is not only to compare currently available macroscopic approaches but above-all to encourage the development of novel next-generation approaches that focus on more microscopic processes. Thus, the benchmark problem includes both macroscopic and microscopic measurement data. In this context, the sub-channel grade void fraction data are regarded as the macroscopic data and the digitized computer graphic images are the microscopic data. The technical topics to be addressed at the workshop include: - Review of the benchmark activities after the 4. Workshop; - Presentation and discussion of summary of comparisons of final submitted results for Exercise 1 of Phase I (I-1); for Exercise 0 of Phase II (II-0); and for Exercise 1 of Phase II (II-1); - Presentation and discussion of comparison of final submitted results for Exercise 2 of Phase I (I-2); - Presentation and discussion of comparison of final submitted results for Exercise 3 of Phase I (I-3); - Presentation and discussion of comparison of final submitted results for Exercise 2 of Phase II (II-2); - Presentation and discussion of preliminary uncertainty results for Exercise 4 of Phase I (I-4); - Presentation and discussion of preliminary uncertainty results for Exercise 3 of Phase II (II-3); - Preparing a special issue in a journal with participants' BFBT papers; - Defining a work plan and schedule outlining actions to advance the two phases of the benchmark activities

  3. Requirements and analysis of electromagnetic compatibility of safety-related instrumentation and control system in nuclear power plants

    Liu Sujuan

    2002-01-01

    The state-of-the-art instrumentation and control system and the influence of their application to the electromagnetic compatibility is analyzed. Based on the present situation of nuclear safety in China and relevant experiences from other countries, the author tries to probe into the requirements and test methods about how safety-related instrument and control system to accommodate electromagnetic interference, radio-frequency interference and power surges in the environments of nuclear power plant so as to develop Chinese safety standards

  4. Contribution to a quantitative assessment model for reliability-based metrics of electronic and programmable safety-related functions

    Hamidi, K.

    2005-10-01

    The use of fault-tolerant EP architectures has induced growing constraints, whose influence on reliability-based performance metrics is no more negligible. To face up the growing influence of simultaneous failure, this thesis proposes, for safety-related functions, a new-trend assessment method of reliability, based on a better taking into account of time-aspect. This report introduces the concept of information and uses it to interpret the failure modes of safety-related function as the direct result of the initiation and propagation of erroneous information until the actuator-level. The main idea is to distinguish the apparition and disappearance of erroneous states, which could be defined as intrinsically dependent of HW-characteristic and maintenance policies, and their possible activation, constrained through architectural choices, leading to the failure of safety-related function. This approach is based on a low level on deterministic SED models of the architecture and use non homogeneous Markov chains to depict the time-evolution of probabilities of errors. (author)

  5. Computerized reactor protection and safety related systems in nuclear power plants. Proceedings of a specialists' meeting. Working material

    1998-01-01

    Though the majority of existing control and protection systems in nuclear power plants use old analogue technology and design philosophy, the use of computers in safety and safety related systems is becoming a current practice. The Specialists Meeting on ''Computerized Reactor Protection and Safety Related Systems in Nuclear Power Plants'' was organized by IAEA (jointly by the Division of Nuclear Power and the Fuel Cycle and the Division of Nuclear Installation Safety), in co-operation with Paks Nuclear Power Plant in Hungary and was held from 27-29 October 1997 in Budapest, Hungary. The meeting focused on computerized safety systems under refurbishment, software reliability issues, licensing experiences and experiences in implemented computerized safety and safety related systems. Within a meeting programme a technical visit to Paks NPP was organized. The objective of the meeting was to provide an international forum for the presentation and discussion on R and D, in-plant experiences in I and C important to safety, backfits and arguments for and reservations against the digital safety systems. The meeting was attended by 70 participants from 16 countries representing NPPs and utility organizations, design/engineering, research and development, and regulatory organizations. In the course of 4 sessions 25 technical presentations were made. The present volume contains the papers presented by national delegates and the conclusions drawn from the final general discussion

  6. Criteria for safety-related nuclear-power-plant operator actions: 1982 pressurized-water-reactor (PWR) simulator exercises

    Crowe, D.S.; Beare, A.N.; Kozinsky, E.J.; Haas, P.M.

    1983-06-01

    The primary objective of the Safety-Related Operator Action (SROA) Program at Oak Ridge National Laboratory is to provide a data base to support development of criteria for safety-related actions by nuclear power plant operators. When compared to field data collected on similar events, a base of operator performance data developed from the simulator experiments can then be used to establish safety-related operator action design evaluation criteria, evaluate the effects of performance shaping factors, and support safety/risk assessment analyses. This report presents data obtained from refresher training exercises conducted in a pressurized water reactor (PWR) power plant control room simulator. The 14 exercises were performed by 24 teams of licensed operators from one utility, and operator performance was recorded by an automatic Performance Measurement System. Data tapes were analyzed to extract operator response times (RTs) and error rate information. Demographic and subjective data were collected by means of brief questionnaires and analyzed in an attempt to evaluate the effects of selected performance shaping factors on operator performance

  7. Gradual linguistic summaries

    Wilbik, A.M.; Kaymak, U.; Laurent, A.; Strauss, O.; Bouchon-Meunier, xx

    2014-01-01

    In this paper we propose a new type of protoform-based linguistic summary – the gradual summary. This new type of summaries aims in capturing the change over some time span. Such summaries can be useful in many domains, for instance in economics, e.g., "prices of X are getting smaller" in eldercare,

  8. The (safety-related) heat exchangers aging management guideline for commercial nuclear power plants, and developments since 1994

    Clauss, J.M.

    1998-01-01

    The US Department of Energy (DOE), in cooperation with the Electric Power Research Institute (EPRI) and US nuclear power plant utilities, is preparing a series of aging management guidelines (AMGs) for commodity types of components (e.g., heat exchangers, electrical cable and terminations, pumps). Commodities are included in this series based on their importance to continued nuclear plant operation and license renewal. The AMGs contain a detailed summary of operating history, stressors, aging mechanisms, and various types of maintenance and surveillance practices that can be combined to create an effective aging management program. Each AMG is intended for use by the systems engineers and plant maintenance staff (i.e., an AMG is intended to be a hands-on technical document rather than a licensing document). The heat exchangers AMG, published in June 1994, includes the following information of interest to nondestructive examination (NDE) personnel: aging mechanisms determined to be non-significant for all applications; aging mechanisms determined to be significant for some applications; effective conventional programs for managing aging; and effective unconventional programs for managing aging. Since the AMG on heat exchangers was published four years ago, a brief review has been conducted to identify emerging regulatory issues, if any. The results of this review and lessons learned from the collective set of AMGs are presented

  9. Executive summary

    2010-01-01

    The special session on Fuel cycle strategies and transition scenarios comprised three invited papers and five oral presentations: INL (USA) was invited to present US activities on fuel cycle transition scenarios; JAEA (Japan) presented the current status of the Japanese nuclear fuel cycle; CEA and EDF (France) gave a presentation on the French fuel cycle strategy and transition scenarios; CEA and INL presented the latest outcomes from the NEA activity on fuel cycle transition scenarios and the European approach; JAEA reported on the results of global scenarios for fast reactor deployment; AECL (Canada) discussed actinide transmutation in Candu reactors, which may efficiently transmute TRU; Materials assessments, which could be used in advanced nuclear fuel cycles from a safeguard perspective, were presented by LANL (USA); The technical session about Impact on P and T on waste management and geological disposal comprised one invited paper and two oral presentations: JAEA presented on the concept of waste management and geological disposal incorporating P and T technology; CIEMAT (Spain) gave a summary of the RED-IMPACT study, which is the study of the impact of P and T on the HLW management programme of the EC; Chalmers University(Sweden) presented an estimation of maximum permissible step losses in P and T processing; The technical session about Progress in transmutation fuels and targets comprised one invited paper and seven oral presentations: The invited talk of the session was given by ITU (EC), on advanced fuel fabrication processes for transmutation; INL presented the development status of transuranic-bearing metal fuels in the USA; CEA summarised European projects on design, development and qualification of advanced fuels for an industrial ADS prototype; The Japanese study of the microstructural evolution and Am migration behaviour in Am-containing MOX fuels at the initial stage of irradiation was presented by JAEA discussed Japanese status on the

  10. Activities in Support of Continuing the Service of Nuclear Power Plant Safety-Related Concrete Structures

    Naus, Dan J.

    2010-01-01

    Nuclear power plant concrete structures are described. In-service inspection and testing requirements in the U.S. are summarized. The license renewal process in the U.S. is outlined and its current status provided. Operating experience related to performance of the concrete structures is presented. Basic components of a program to manage aging of the concrete structures are identified and described: degradation mechanisms, damage models, and material performance; assessment and remediation (i.e., component selection, in-service inspection, non-destructive examinations, and remedial actions); and estimation of performance at present or some future point in time (i.e., application of structural reliability theory to the design and optimization of in-service inspection/maintenance strategies, and determination of the effects of degradation on plant risk). Finally, areas are noted where additional research would be of benefit to aging management of nuclear power plant concrete structures.

  11. Reliability Analysis on NPP's Safety-Related Control Module with Field Data

    Lee, Sang Yong; Jung, Jae Hyun; Kim, Seong Hun

    2006-01-01

    The automatic control systems used in nuclear power plant (NPP) consists of numerous control modules that can be considered to be a network of components various complex ways. The control modules require relatively high reliability than industrial electronic products. Reliability prediction provides the rational basis of system designs and also provides the safety significance of system operations. The aim of this paper is to minimize the deficiencies of the traditional reliability prediction method calculation using the available field return data. This way is possible to do more realistic reliability assessment. SAMCHANG Enterprise Company (SEC) has established database containing high quality data at the module and component level from module maintenance in NPP. On the basis of these, this paper compares results that add failure record (field data) to Telcordia-SR-332 reliability prediction model with MIL-HDBK-217F prediction results

  12. Executive Summary

    2012-01-01

    The OECD Nuclear Energy Agency (NEA) Integration Group for the Safety Case (IGSC) organised a workshop to assess current understanding on the use of cementitious materials in radioactive waste disposal. The workshop was hosted by the Belgian Agency for Radioactive Waste and Enriched Fissile Materials (Ondraf/Niras), in Brussels, Belgium on 17-19 November 2009. The workshop brought together a wide range of people involved in supporting safety case development and having an interest in cementitious materials: namely, cement and concrete experts, repository designers, scientists, safety assessors, disposal programme managers and regulators. The workshop was designed primarily to consider issues relevant to the post-closure safety of radioactive waste disposal, but also addressed some related operational issues, such as cementitious barrier emplacement. Where relevant, information on cementitious materials from analogous natural and anthropogenic systems was also considered. The workshop agenda is included as Appendix A. The workshop focused on: - The uses of different cementitious materials in various repository designs. - The evolution of cementitious materials over long time scales (1000s to 100000s of years). - The interaction of cementitious materials with surrounding components of the repository (e.g. waste, container, buffer, backfill, host rock). - The workshop comprised: - Plenary sessions in which the state-of-the-art on repository design and understanding the phenomenology of cementitious materials and their interactions were presented and discussed. - Dedicated working group sessions, which were used to discuss key safety assessment and safety case questions in more detail. For example: How strong is the scientific basis for incorporating the various aspects of the behaviour and interactions of cementitious materials in safety assessments and safety cases? How can the behaviour and interactions of cementitious materials best be incorporated within the

  13. Executive summary

    2015-01-01

    Since 2007, the OECD Nuclear Energy Agency has been organising a series of workshops on Structural Materials for Innovative Nuclear Systems. The third meeting was held on 7-9 October 2013 in Idaho Falls (United States). The main objectives of this workshop are to stimulate an exchange of information on current materials R and D programmes for different innovative nuclear systems. The main topics of the workshop covered fundamental studies, modelling and experiments on innovative structural materials including cladding materials for the range of advanced nuclear systems such as thermal/fast systems, sub-critical systems, as well as fusion systems. During the workshop, the following topics were discussed: - Fundamental studies; - Metallic materials; - Ceramic materials; - Novel materials pathways; - Ion vs neutron irradiation. Fundamental studies focused on the identification of mechanisms driving the response of materials under the conditions expected in innovative nuclear systems. These mechanisms may have acted at the atomic or higher scale with the application of multi-scale approaches, together with related problems of scale-bridging or numerical methods, were of special interest. Moreover, irradiation experiments and subsequent characterisation of materials with analytical techniques were included in the session if aimed at better understanding the acting mechanisms or drawing physics-based correlations. Metal alloys, ceramic and ceramic composites included in- and out-of-core applications which took into account the scope of: data availability and gaps (considering also licensing issues); experimental and modelling needs for specific components or degradation modes; the link between R and D, standardisation and experimental protocols; coolant effects and mechanical properties. Code development and implementation plans were also discussed. Application of SiC composites to LWR systems was of interest as an advanced concept. Novel materials pathways considered

  14. Monitoring of operational reliability of safety-related I and C subsystems at the Dukovany NPP

    Fuchs, P.; Sagl, P.; Zlamal, P.

    2007-01-01

    First, the situation existing in the data base in 1999, i.e. before the monitoring and the operational reliability monitoring concept were introduced, is highlighted. The technique of data processing is described with focus on the assessment of the relevancy of the records, component failure rate monitoring, estimation of basic statistical parameters, evaluation of the feasibility of component failure (or failure latency) detection, assessment of the mean time to repair, FMEA of the basic components (relays end measuring chains) to establish spurious signals and dangerous failure ratio. The reliability assessment of the system functions is based on structural reliability calculations (common cause failures not included). The outcomes from the operational reliability monitoring are presented in the form of a representative set of data, graphic charts and results of system function reliability assessment. Prospects for upgrading the I and C operational reliability monitoring system to the benefit of NPP Dukovany operating economy (life cycle costs evaluation, spare parts planning, RCM application) are outlined. (author)

  15. Seismic performance assessment of base-isolated safety-related nuclear structures

    Huang, Y.-N.; Whittaker, A.S.; Luco, N.

    2010-01-01

    Seismic or base isolation is a proven technology for reducing the effects of earthquake shaking on buildings, bridges and infrastructure. The benefit of base isolation has been presented in terms of reduced accelerations and drifts on superstructure components but never quantified in terms of either a percentage reduction in seismic loss (or percentage increase in safety) or the probability of an unacceptable performance. Herein, we quantify the benefits of base isolation in terms of increased safety (or smaller loss) by comparing the safety of a sample conventional and base-isolated nuclear power plant (NPP) located in the Eastern U.S. Scenario- and time-based assessments are performed using a new methodology. Three base isolation systems are considered, namely, (1) Friction Pendulum??? bearings, (2) lead-rubber bearings and (3) low-damping rubber bearings together with linear viscous dampers. Unacceptable performance is defined by the failure of key secondary systems because these systems represent much of the investment in a new build power plant and ensure the safe operation of the plant. For the scenario-based assessments, the probability of unacceptable performance is computed for an earthquake with a magnitude of 5.3 at a distance 7.5 km from the plant. For the time-based assessments, the annual frequency of unacceptable performance is computed considering all potential earthquakes that may occur. For both assessments, the implementation of base isolation reduces the probability of unacceptable performance by approximately four orders of magnitude for the same NPP superstructure and secondary systems. The increase in NPP construction cost associated with the installation of seismic isolators can be offset by substantially reducing the required seismic strength of secondary components and systems and potentially eliminating the need to seismically qualify many secondary components and systems. ?? 2010 John Wiley & Sons, Ltd.

  16. Technical basis for evaluating electromagnetic and radio-frequency interference in safety-related I ampersand C systems

    Ewing, P.D.; Korsah, K.

    1994-04-01

    This report discusses the development of the technical basis for the control of upsets and malfunctions in safety-related instrumentation and control (I ampersand C) systems caused by electromagnetic and radio-frequency interference (EMI/RFI) and power surges. The research was performed at the Oak Ridge National Laboratory (ORNL) and was sponsored by the USNRC Office of Nuclear Regulatory Research (RES). The motivation for research stems from the safety-related issues that need to be addressed with the application of advanced I ampersand C systems to nuclear power plants. Development of the technical basis centered around establishing good engineering practices to ensure that sufficient levels of electromagnetic compatibility (EMC) are maintained between the nuclear power plant's electronic and electromechanical systems known to be the source(s) of EMI/RFI and power surges. First, good EMC design and installation practices need to be established to control the impact of interference sources on nearby circuits and systems. These EMC good practices include circuit layouts, terminations, filtering, grounding, bonding, shielding, and adequate physical separation. Second, an EMI/RFI test and evaluation program needs to be established to outline the tests to be performed, the associated test methods to be followed, and carefully formulated acceptance criteria based on the intended environment to ensure that the circuit or system under test meets the recommended guidelines. Third, a program needs to be developed to perform confirmatory tests and evaluate the surge withstand capability (SWC) and of I ampersand C equipment connected to or installed in the vicinity of power circuits within the nuclear power plant. By following these three steps, the design and operability of safety-related I ampersand C systems against EMI/RFI and power surges can be evaluated, acceptance criteria can be developed, and appropriate regulatory guidance can be provided

  17. Major results from safety-related integral effect tests with VISTA-ITL for the SMART design

    Park, H. S.; Min, B. Y.; Shin, Y. C.; Yi, S. J.

    2012-01-01

    A series of integral effect tests (IETs) was performed by the Korea Atomic Energy Research Inst. (KAERI) using the VISTA integral test loop (VISTA-ITL) as a small-scale IET program. Among them this paper presents major results acquired from the safety-related IETs with the VISTA-ITL facility for the SMART design. Three small-break loss-of-coolant accident (SBLOCA) tests of safety injection system (SIS) line break, shutdown cooling system (SCS) line break and pressurizer safety valve (PSV) line break were successfully performed and the transient characteristics of a complete loss of flowrate (CLOF) was simulated properly with the VISTA-ITL facility. (authors)

  18. Modification and backfitting at the Barsebaeck Nuclear Power Plant Unit 1 and 2 in safety related systems

    Karlsson, Leif; Nilsson, Ove; Lidh, B.

    1995-05-01

    This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Barsebaeck, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 3 refs

  19. Establishing management information system to solve the information management problem of nuclear safety related personnel's qualification management

    Sun Haipeng; Liu Zhijun; Li Tianshu

    2013-01-01

    With the rapid progress of nuclear energy and nuclear technology utilization, nuclear safety related personnel play an increasingly important role in ensuring nuclear safety. NNSA personnel qualification management information system conducts a multi-faceted, effective, real-time monitoring and information collection for nuclear safety staff practice unit management, knowledge management, license application, appraisal management or supervision, training management or supervision and certified staff management, and also is a milestone for NNSA to build the state department with 'five-feature' (learning-oriented, service-oriented, economical, innovative, clean-type). (authors)

  20. Application of the Commercial Grade Item (CGI) Dedication Process for Procurement of Nuclear Safety Related Items at Nuclear Power Plant Krsko (NEK)

    Heruc, Z.; Pozar, J.

    1998-01-01

    CGI procurement is a process whereby parts are brought without imposing Appendix B Quality Assurance requirements on the supplier, and than dedicated for use in safety-related applications. The dedication process involves 1) based upon required safety function, an engineering evaluation to identify critical characteristic of the item and specification of acceptance criteria; and 2) quality control activities to ensure the item(s) supplied meets the acceptance criteria specified. CGI Dedication supports the supply of certified components/parts for the plant operation in an environment where the number of nuclear qualified suppliers diminishes. It requires a more active role of the plant personnel, therefore presenting an additional burden on human resources, but at the same time increases the technical KNOW-HOW and improves the confidence of test and inspection data presented in the certificates. Very often it is also cost beneficial. This paper is a continuation to last year presentation of the introduction of this method into NEK's procurement process and presents the current approach and some practical examples. (author)

  1. NRC Information Notice No. 92-27: Thermally induced accelerated aging and failure of ITE/Gould a.c. relays used in safety-related applications

    Rossi, C.E.

    1993-01-01

    On November 23, 1991, while performing an eighteen month engineered safety features operability test, the licensee for the Millstone Nuclear Power Station, Unit 3, noted that control power was interrupted to three safety-related motor operated valves (MOVs). The valves were located in the charging, component cooling water, and steam generator atmospheric dump systems. The licensee inspected the valves' control power circuitry and determined that three normally energized auxiliary relays had failed. These relays provided control power alarms and thermal overload protection for the MOVs. The relay failures rendered each valve inoperable. The relays, which had been in service for about seven years, were class J10 relays with J20M magnet block assemblies and standard G10JA126, 120V, 60 cycle coil assemblies manufactured by the ITE/Gould Manufacturing Company. Inspection of the relays revealed that the movable plastic armature carrier, which surrounds the core and coil, and the retainer for the magnet yoke assembly were discolored, brittle and severely cracked. Insulation degradation was severe, allowing electrical shorts to develop within the coils. The licensee concluded that the failures resulted from the thermal aging of the coil assemblies and plastic parts near the coil assemblies

  2. Regulatory surveillance of safety related maintenance at nuclear power plants. Report of a technical committee meeting

    1997-08-01

    The operational safety and reliability of a nuclear power plant as well as its availability for electricity generation depend on, among other things, its maintenance programme. Regulatory bodies therefore have considerable interest in maintenance activities. There are several approaches to maintenance, i.e. reliability centered maintenance or risk focused maintenance, aimed at optimizing maintenance by focusing on important components or systems. These approaches may result in significant changes to maintenance activities and therefore have to be considered for regulatory acceptance. In order to review and discuss the status of maintenance regulation in participating countries, the IAEA convened a Technical Committee Meeting on Regulatory Oversight of Maintenance Activities at Nuclear Power Plants in Vienna from 9 to 13 October 1995. The meeting was attended by 16 experts from 11 countries. In addition to the consideration of papers that were presented and which are reproduced here, extensive group and panel discussions took place during the meeting. These covered three main topics: general features and basic characteristics of maintenance regulation, regulatory acceptance of maintenance optimization and use of PSA for maintenance optimization. The discussion are summarized in Section 2. Section 3 discusses the following three additional topics: regulatory involvement in the maintenance programme, modifications to the maintenance programme and personnel related aspects of maintenance. The conclusions are presented in Section 4. Figs, tabs

  3. Significance of Alkali-Silica reaction in nuclear safety-related concrete structures

    Le Pape, Y.; Field, K.G.; Mattus, C.H.; Naus, D.J.; Busby, J.T.; Saouma, V.; Ma, Z.J.; Cabage, J.V.; Guimaraes, M.

    2015-01-01

    Nuclear Power Plant license renewal up to 60 years and possible life extension beyond has established a renewed focus on long-term aging of nuclear generating stations materials, and particularly, on concrete. Large irreplaceable sections of most nuclear generating stations include concrete components. The Expanded Materials Degradation Analysis, jointly performed by the Department of Energy, the U.S. Nuclear Regulatory Commission, the Academia and the Power Generation Industry, identified the need to develop a consistent knowledge base of alkali-silica reaction (ASR) within concrete as an urgent priority (Graves et al., 2014). ASR results in an expansion of Concrete produced by the reaction between alkali (generally from cement), reactive aggregate (like amorphous silica) and water absorption. ASR causes expansion, cracking and loss of mechanical properties. Considering that US commercial reactors in operation enter the age when ASR distress can be potentially observed and that numerous non-nuclear infrastructures (transportation, energy production) in a majority of the States have already experienced ASR-related concrete degradation, the susceptibility and significance of ASR for nuclear concrete structures must be addressed. This paper outlines an on-going research program including the investigation of the possibility of ASR in nuclear power plants, and the assessment of the residual shear bearing capacity of ASR-subjected nuclear structures. (authors)

  4. Spent fuel reprocessing and minor actinide partitioning safety related research at the UK National Nuclear Laboratory

    Carrott, Michael; Flint, Lauren; Gregson, Colin; Griffiths, Tamara; Hodgson, Zara; Maher, Chris; Mason, Chris; McLachlan, Fiona; Orr, Robin; Reilly, Stacey; Rhodes, Chris; Sarsfield, Mark; Sims, Howard; Shepherd, Daniel; Taylor, Robin; Webb, Kevin; Woodall, Sean; Woodhead, David

    2015-01-01

    The development of advanced separation processes for spent nuclear fuel reprocessing and minor actinide recycling is an essential component of international R and D programmes aimed at closing the nuclear fuel cycle around the middle of this century. While both aqueous and pyrochemical processes are under consideration internationally, neither option will gain broad acceptance without significant advances in process safety, waste minimisation, environmental impact and proliferation resistance; at least when compared to current reprocessing technologies. The UK National Nuclear Laboratory (NNL) is developing flowsheets for innovative aqueous separation processes. These include advanced PUREX options (i.e. processes using tributyl phosphate as the extractant for uranium, plutonium and possibly neptunium recovery) and GANEX (grouped actinide extraction) type processes that use diglycolamide based extractants to co-extract all transuranic actinides. At NNL, development of the flowsheets is closely linked to research on process safety, since this is essential for assessing prospects for future industrialisation and deployment. Within this context, NNL is part of European 7. Framework projects 'ASGARD' and 'SACSESS'. Key topics under investigation include: hydrogen generation from aqueous and solvent phases; decomposition of aqueous phase ligands used in separations prior to product finishing and recycle of nitric acid; dissolution of carbide fuels including management of organics generated. Additionally, there is a strong focus on use of predictive process modelling to assess flowsheet sensitivities as well as engineering design and global hazard assessment of these new processes. (authors)

  5. Towards the Development of a Methodology for the Cyber Security Analysis of Safety Related Nuclear Digital I and C Systems

    Khand, Parvaiz Ahmed; Seong, Poong Hyun

    2007-01-01

    In nuclear power plants the redundant safety related systems are designed to take automatic action to prevent and mitigate accident conditions if the operators and the non-safety systems fail to maintain the plant within normal operating conditions. In case of an event, the failure of these systems has catastrophic consequences. The tendency in the industry over the past 10 years has been to use of commercial of the shelf (COTS) technologies in these systems. COTS software was written with attention to function and performance rather than security. COTS hardware usually designed to fail safe, but security vulnerabilities could be exploited by an attacker to disable the fail safe mechanisms. Moreover, the use of open protocols and operating systems in these technologies make the plants to become vulnerable to a host of cyber attacks. An effective security analysis process is required during all life cycle phases of these systems in order to ensure the security from cyber attacks. We are developing a methodology for the cyber security analysis of safety related nuclear digital I and C Systems. This methodology will cover all phases of development, operation and maintenance processes of software life cycle. In this paper, we will present a security analysis process for the concept stage of software development life cycle

  6. Summary report for the second TUV-workshop proceedings on living PSA application

    1991-01-01

    This workshop on living PSA Application was organized to support the OECD/NEA CSNI-Principal Working Group No.5 on Risk Assessment for an international exchange of experience on living PSA application. The first session was devoted to Living PSA Applications and the second session to Tools for Living PSA. Living PSA Applications: Reasons for performing PSA (regulatory requirement, targets; corporate requirement, targets; safety related activity prioritization; other); Logistic of Living PSA Management (Corporate management involvement, Decision making levels and guidance, Plant level involvement, Required personnel commitment, Frequency and extent of re-quantification of PSA, Types of safety/risk parameters to be monitored, Quality assurance on maintaining Living PSA); Examples of Application (Experiences of application, State of Living PSA/e.g. all accident sequences involved, Details of component level involvement). Tools for Living PSA: Data Collection Systems and Codes (Source and type of data collected, Probabilistic parameter quantification, Interface to basic event data, Data code systems). An executive summary of the workshop is given

  7. The Development of the Safety Related Valve Class 1E Electrical Motor, the Target and the Results

    Saban, I.; Grgic, D.; Fancev, T.; Flegar, Lj.; Novosel, N.

    1996-01-01

    The development of the safety related valves class 1E electric motor is described. The design implemented in order to satisfy the 1E requirements, and a way in which related 1E standards are addressed, are shown. The development was realized in three stages. In the first stage eight motorettes were made and the insulation system was tested. In the second stage the motor was produced in accordance with producer's prototype QA program. In the third stage part of the testing of the produced motor was made. The results of the testing, finished until now, show that produced motor, as well as similarly produced motors, is able to perform its safety function in the design bases accident conditions as requested by class 1E requirements. The rest of the testing (LOCA test) can be made on the same or similar motor in the future. (author)

  8. Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

    Korsah, K.; Wood, R.T.; Hassan, M.; Tanaka, T.J.

    1998-01-01

    This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

  9. Impact of state-of-the-art instrumentation on safety-related experimental studies proposed in containment studies facility (CSF)

    Gole, N.V.; Markandeya, S.G.; Subramaniam, K.; Ghosh, A.K.

    2002-01-01

    Full text: Conducting an experimental program for safety related studies for nuclear power plants (NPPs) is an extremely laborious and time-consuming task due to several reasons. Requirement for frequent replacements, testing and recalibration of a large number of instruments is one of them. Off-line analysis leading to identification of errors is another. A particular test may have to be abandoned based on such analysis. Following the rapid advances in instrumentation, a larger number of options are now available, which make experimentation easy. CSF is one of the upcoming facilities wherein deployment of state-of-the art became inevitable. This paper discusses in detail the design intent of instrumentation, the state-of-the-art instrumentation provisions made to fulfill it the overall impact of this on successful experimentation

  10. Review of safety related control room function research based on experience from nuclear power plants in Finland

    Juslin, K.; Wahlstroem, B.; Rinttilae, E.

    1985-01-01

    A comprehensive human engineering research programme was established in the second half of the 1970's at the Technical Research Centre of Finland (VTT). The research is performed in cooperation with the utility companies Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (TVO) and includes topics such as Handling of alarm information, Disturbance analysis systems, Assessment of control rooms and Validation of safety parameter display systems. Reference is also made to the Finnish contribution to the OECD Halden Reactor Project (Halden) and the Nordic Liaison Committee for Atomic Energy (NKA) research projects. In this paper feasible realization alternatives of safety related control room functions are discussed on the basis of experience from the nuclear power plants in Finland, which at present are equipped with extensive process computer systems. A proposal for future power plant information systems is described. It is intended that this proposal will serve as the basis for future computer systems at nuclear power plants in Finland. (author)

  11. Undetected latent failures of safety-related systems. Preliminary survey of events in nuclear power plants 1980-1997

    Lydell, B.

    1998-03-01

    This report summarizes results and insights from a preliminary survey of events involving undetected, latent failures of safety-related systems. The survey was limited to events where mispositioned equipment (e.g., valves, switches) remained undetected, thus rendering standby equipment or systems unavailable for short or long time periods. Typically, these events were symptoms of underlying latent errors (e.g., design errors, procedure errors, unanalyzed safety conditions) and programmatic errors. The preliminary survey identified well over 300 events. Of these, 95 events are documented in this report. Events involving mispositioned equipment are commonplace. Most events are discovered soon after occurrence, however. But as evidenced by the survey results, some events remained undetected beyond several shift changes. The recommendations developed by the survey emphasize the importance of applying modern root cause analysis techniques to the event analysis to ensure that the causes and implications of occurred events are fully understood

  12. Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

    Korsah, K.; Wood, R.T. [Oak Ridge National Lab., TN (United States); Hassan, M. [Brookhaven National Lab., Upton, NY (United States); Tanaka, T.J. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01

    This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.

  13. Conceptual design of an integrated information system for safety related analysis of nuclear power plants (IRIS Phase 1)

    Hofer, K.; Zehnder, P.; Galperin, A.

    1994-01-01

    This report deals with a conceptual design of an integrated information management system, called PSI-IRIS, as needed to assist the analysts for reactor safety related investigations on Swiss nuclear power plants within the project STARS. Performing complicated engineering analyses of an NPP requires storage and manipulation of a large amount of information, both data and knowledge. This information is characterized by its multi-disciplinary nature, complexity, and diversity. The problems caused by inefficient and lengthy manual operations involving the data flow management within the framework of the safety related analysis of an NPP, can be solved by applying computer aided engineering (CAE) principles. These principles are the basis for the design of the integrated information management system PSI-IRIS presented in this report. The basic idea is to create a computerized environment, which includes both database and functional capabilities. The database of the PSI-IRIS consists of two parts, an NPP generic database (GDB) and a collection of analysis results (CASE L IB). The GDB includes all technical plant data and information needed to generate input decks for all computer codes utilized within the STARS project. The CASE L IB storage contains the accumulated knowledge, input decks, and result files of the NPP transient analyses. Considerations and analysis of the data types and the required data manipulation capabilities as well as operational requirements resulted in the choice of an object-oriented database management system (OODBMS) as a development platform for solving the software engineering problems. Several advantages of OODBMS's over conventional relational database management systems were found of crucial importance, especially providing the necessary flexibility for different data types and the potential for extensibility. (author) 15 figs., tabs., 20 refs

  14. Other components

    Anon.

    1993-01-01

    This chapter includes descriptions of electronic and mechanical components which do not merit a chapter to themselves. Other hardware requires mention because of particularly high tolerance or intolerance of exposure to radiation. A more systematic analysis of radiation responses of structures which are definable by material was given in section 3.8. The components discussed here are field effect transistors, transducers, temperature sensors, magnetic components, superconductors, mechanical sensors, and miscellaneous electronic components

  15. FEMA Disaster Declarations Summary

    Department of Homeland Security — The FEMA Disaster Declarations Summary is a summarized dataset describing all federally declared disasters, starting with the first disaster declaration in 1953,...

  16. Working Safely at Some Times and Unsafely at Others: A Typology and Within-Person Process Model of Safety-Related Work Behaviors.

    Beus, Jeremy M; Taylor, William D

    2017-06-22

    Why do individuals choose to work safely in some instances and unsafely in others? Though this inherently within-person question is straightforward, the preponderance of between-person theory and research in the workplace safety literature is not equipped to answer it. Additionally, the limited way in which safety-related behaviors tend to be conceptualized further restricts understanding of why individuals vary in their safety-related actions. We use a goal-focused approach to conceptually address this question of behavioral variability and contribute to workplace safety research in 2 key ways. First, we establish an updated typology of safety-related behaviors that differentiates behaviors based on goal choice (i.e., safe vs. unsafe behaviors), goal-directedness (i.e., intentional vs. unintentional behaviors), and the means of goal pursuit (i.e., commission vs. omission and promotion vs. prevention-focused behaviors). Second, using an expectancy-value theoretical framework to explain variance in goal choice, we establish within-person propositions stating that safety-related goal choice and subsequent behaviors are a function of the target of safety-related behaviors, the instrumentality and resource requirement of behaviors, and the perceived severity, likelihood, and immediacy of the threats associated with behaviors. Taken together, we define what safety-related behaviors are, explain how they differ, and offer propositions concerning when and why they may vary within-persons. We explore potential between-person moderators of our theoretical propositions and discuss the practical implications of our typology and process model of safety-related behavior. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  17. Detecting decay in wood components

    R.J. Ross; X. Wang; B.K. Brashaw

    2005-01-01

    This chapter presents a summary of the Wood and Timber Condition Assessment Manual. It focuses on current inspection techniques for decay detection and provides guidelines on the use of various non-destructive evaluation (NDE) methods in locating and defining areas of deterioration in timber bridge components and other civil structures.

  18. Summary information report

    1982-07-01

    The Summary Information Report (SIR) provides summary data concerning NRC and its licensees for general use by the Chairman, other Commissioners and Commission staff offices, the Executive Director for Operations, and the Office Directors. SIR is published quarterly by the Management Information Branch (49-27834) of the Office of Resource Management

  19. Electronic components

    Colwell, Morris A

    1976-01-01

    Electronic Components provides a basic grounding in the practical aspects of using and selecting electronics components. The book describes the basic requirements needed to start practical work on electronic equipment, resistors and potentiometers, capacitance, and inductors and transformers. The text discusses semiconductor devices such as diodes, thyristors and triacs, transistors and heat sinks, logic and linear integrated circuits (I.C.s) and electromechanical devices. Common abbreviations applied to components are provided. Constructors and electronics engineers will find the book useful

  20. Model Adoption Exchange Payment System: Executive Summary.

    Ambrosino, Robert J.

    This executive summary provides a brief description of the Model Adoption Exchange Payment System (MAEPS), a unique payment system aimed at improving the delivery of adoption exchange services throughout the United States. Following a brief introductory overview, MAEPS is described in terms of (1) its six components (registration, listing,…

  1. Seismic simulation and functional performance evaluation of a safety related, seismic category I control room emergency air cleaning system

    Manley, D.K.; Porco, R.D.; Choi, S.H.

    1985-01-01

    Under a nuclear contract MSA was required to design, manufacture, seismically test and functionally test a complete Safety Related, Seismic Category I, Control Room Emergency Air Cleaning System before shipment to the Yankee Atomic Electric Company, Yankee Nuclear Station in Rowe, Massachusetts. The installation of this system was required to satisfy the NRC requirements of NUREG-0737, Section III, D.3.4, ''Control Room Habitability''. The filter system tested was approximately 3 ft. wide by 8 ft. high by 18 ft. long and weighed an estimated 8300 pounds. It had a design flow rate of 3000 SCFM and contained four stages of filtration - prefilters, upstream and downstream HEPA filters and Type II sideload charcoal adsorber cells. The filter train design followed the guidelines set forth by ANSI/ASME N509-1980. Seismic Category I Qualification Testing consisted of resonance search testing and triaxial random multifrequency testing. In addition to ANSI/ASME N510-1980 testing, triaxial response accelerometers were placed at specific locations on designated prefilters, HEPA filters, charcoal adsorbers and test canisters along with accelerometers at the corresponding filter seal face locations. The purpose of this test was to demonstrate the integrity of the filters, filter seals, and monitor seismic response levels which is directly related to the system's ability to function during a seismic occurrence. The Control Room Emergency Air Cleaning System demonstrated the ability to withstand the maximum postulated earthquake for the plant site by remaining structurally sound and functional

  2. A Study of Time Response for Safety-Related Operator Actions in Non-LOCA Safety Analysis

    Lee, Min Seok; Lee, Sang Seob; Park, Min Soo; Lee, Gyu Cheon; Kim, Shin Whan [KEPCO E and C Company, Daejeon (Korea, Republic of)

    2014-10-15

    The classification of initiating events for safety analysis report (SAR) chapter 15 is categorized into moderate frequency events (MF), infrequent events (IF), and limiting faults (LF) depending on the frequency of its occurrence. For the non-LOCA safety analysis with the purpose to get construction or operation license, however, it is assumed that the operator response action to mitigate the events starts at 30 minutes after the initiation of the transient regardless of the event categorization. Such an assumption of corresponding operator response time may have over conservatism with the MF and IF events and results in a decrease in the safety margin compared to its acceptance criteria. In this paper, the plant conditions (PC) are categorized with the definitions in SAR 15 and ANS 51.1. Then, the consequence of response for safety-related operator action time is determined based on the PC in ANSI 58.8. The operator response time for safety analysis regarding PC are reviewed and suggested. The clarifying alarm response procedure would be required for the guideline to reduce the operator response time when the alarms indicate the occurrence of the transient.

  3. Implementation of digital safety related I and C systems at nuclear power plants. A systematic approach to training

    Roedig, Peter; Schoenfelder, Christian

    2012-01-01

    In the past, refurbishment or modernization projects at nuclear power plants (NPP) dealing with the AREVA product for safety related digital instrumentation and control (I and C) systems, i.e. TELEPERM registered XS (TXS), regularly led to the development and implementation of different project specific training courses. They mostly dealt with a basic introduction to TELEPERM registered XS, as well as project specific engineering of TELEPERM registered XS and maintenance of the TELEPERM registered XS system supplied with the project. However, it gradually emerged that diverse training needs of different personnel involved in refurbishment or modernization projects as well as in new build projects had to be considered in more detail. Additionally, each target group, e.g. project managers, project engineers, technical engineers, commissioning engineers, operating and maintenance personnel, will have to work with TELEPERM registered XS at different phases within a project. Consequently, it became necessary to take into account the diverse training and project needs. According to the Systematic Approach to Training (SAT) process as developed and promoted by the International Atomic Energy Agency (IAEA), a job and task analysis was performed. After identification of related training needs and redesigning as well as modification or development of appropriate training material, a comprehensive, standardized TELEPERM registered XS training offer is now available at the AREVA Reactor Training Center. This training offer can be easily adapted to project or customer specific requirements. (orig.)

  4. Investigations of safety-related parameters applying a new multi-group diffusion code for HTR transients

    Kasselmann, S.; Druska, C.; Lauer, A.

    2010-01-01

    The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)

  5. The application of VMEbus system to the safety related parameters indication and alarm system for nuclear power plants

    Lee, Chul Kwon; Koo, In Soo; Jang, Gwi Sook; Shin, Jae Hwal

    1996-12-01

    This report presents the basic feature, the status of technical development, and it`s application for the VMEbus which has been utilized in industrial application such as controller, robotics, automation control. The application software of VMEbus is also reviewed. The design considerations are presented when applying the system to the instrumentation and control technique of nuclear power plants. The conceptual design of safety related parameter using the integrated VMEbus system. The results indicate that the application of VMEbus has advantages such as easy maintenance, accurate and reliable operation, easy expansion and upgrade. Also, the integrated VMEbus system is capable of limited real-time processing because it can be processed by multi-processors and can reduce the effort of software development by using off-the-shelf software. However, the adoption of digital system is produced problems such as software common mode failure, EMI and RFI, and verification and validation methods of off-the-shelf hardware and software. To resolve these problems in the future, further research are required. (author). 7 tabs., 19 figs., 24 refs.

  6. Regulatory analysis for the resolution of generic issue 57: Effects of Fire Protection System Actuation on Safety-Related Equipment

    Woods, H.W.

    1993-10-01

    Actuation of Fire Protection Systems (FPS) in Nuclear Power Plants have resulted in adverse interactions with equipment important to safety. Precursor operational experience has shown that 37% of all FPS actuations damaged some equipment, and 20% of all FPS actuations have resulted in a plant transient and reactor trip. On an average 0.17 FPS actuations per reactor year have been experienced in nuclear power plants in this country. This report presents the regulatory analysis for GI-57, ''Effects of Fire Protection System Actuation on Safety-Related Equipment''. The risk reduction estimates, cost/benefit analyses, and other insights gained during this effort have shown that implementation of the recommendations contained in this report can significantly reduce risk, and that these improvements can be warranted in accordance with the backfit rule, 10 CFR 50.109(a)(3). However, plant specific analyses are required in order to identify such improvements. Generic analyses can not serve to identify improvements that could be warranted for individual, specific plants. Plant specific analyses of the type needed for this purpose are underway as part of the Individual Plant Examination of External Events (IPEEE) program

  7. Undetected latent failures of safety-related systems. Preliminary survey of events in nuclear power plants 1980-1997

    Lydell, B. [RSA Technologies, Vista, CA (United States)

    1998-03-01

    This report summarizes results and insights from a preliminary survey of events involving undetected, latent failures of safety-related systems. The survey was limited to events where mispositioned equipment (e.g., valves, switches) remained undetected, thus rendering standby equipment or systems unavailable for short or long time periods. Typically, these events were symptoms of underlying latent errors (e.g., design errors, procedure errors, unanalyzed safety conditions) and programmatic errors. The preliminary survey identified well over 300 events. Of these, 95 events are documented in this report. Events involving mispositioned equipment are commonplace. Most events are discovered soon after occurrence, however. But as evidenced by the survey results, some events remained undetected beyond several shift changes. The recommendations developed by the survey emphasize the importance of applying modern root cause analysis techniques to the event analysis to ensure that the causes and implications of occurred events are fully understood. 7 refs, 4 tabs, 3 figs. Also available at the SKI Home page: //www.ski.se.

  8. The application of VMEbus system to the safety related parameters indication and alarm system for nuclear power plants

    Lee, Chul Kwon; Koo, In Soo; Jang, Gwi Sook; Shin, Jae Hwal.

    1996-12-01

    This report presents the basic feature, the status of technical development, and it's application for the VMEbus which has been utilized in industrial application such as controller, robotics, automation control. The application software of VMEbus is also reviewed. The design considerations are presented when applying the system to the instrumentation and control technique of nuclear power plants. The conceptual design of safety related parameter using the integrated VMEbus system. The results indicate that the application of VMEbus has advantages such as easy maintenance, accurate and reliable operation, easy expansion and upgrade. Also, the integrated VMEbus system is capable of limited real-time processing because it can be processed by multi-processors and can reduce the effort of software development by using off-the-shelf software. However, the adoption of digital system is produced problems such as software common mode failure, EMI and RFI, and verification and validation methods of off-the-shelf hardware and software. To resolve these problems in the future, further research are required. (author). 7 tabs., 19 figs., 24 refs

  9. Guide on a national system for collecting, assessing and disseminating information on safety-related events in nuclear power plants

    1983-02-01

    There is a wide spectrum of safety significance in the events that can occur during nuclear power plant operations. It is important that lessons be learned from safety-related events (hereinafter referred to as unusual events) so as to improve the safety of nuclear power plants. Hence formal procedures should be established for this purpose. The purpose of this document is to provide guidance to Member States for establishing a system (hereinafter referred to as a national system) for collecting, storing, retrieving, assessing and disseminating information on unusual events in nuclear power plants. The guidance given is based on experience gained in the use of existing national and international systems. This guide covers a national system that is part of a programme to improve nuclear power plant safety using experience gained from operating plants both within and outside the country. Implementing the recommendations in this guide would render any national system compatible with other national systems and facilitate the participation in the IAEA System for Reporting Unusual Events with Safety Significance (hereinafter referred to as the IAEA Incident Reporting System, IAEA-IRS) for more widespread dissemination of lessons learned from nuclear power plant operation

  10. Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-category I systems for Palisades nuclear power plant

    Collins, E.K.

    1979-10-01

    The technical evaluation is presented of Consumers Power Company's Palisades nuclear power plant to determine whether the failure of any non-Category I (seismic) equipment could result in a condition, such as flooding, that might potentially adversely affect the performance of safety-related equipment required for the safe shutdown of the facility or to mitigate the consequences of an accident. Criteria developed by the US Nuclear Regulatory Commission were used to evaluate the acceptability of the existing protection as well as measures taken by Consumers Power Company to minimize the danger of flooding and to protect safety-related equipment

  11. Summary big data

    2014-01-01

    This work offers a summary of Cukier the book: "Big Data: A Revolution That Will Transform How we Live, Work, and Think" by Viktor Mayer-Schonberg and Kenneth. Summary of the ideas in Viktor Mayer-Schonberg's and Kenneth Cukier's book: " Big Data " explains that big data is where we use huge quantities of data to make better predictions based on the fact we identify patters in the data rather than trying to understand the underlying causes in more detail. This summary highlights that big data will be a source of new economic value and innovation in the future. Moreover, it shows that it will

  12. Biofuels: Project summaries

    1994-07-01

    The US DOE, through the Biofuels Systems Division (BSD) is addressing the issues surrounding US vulnerability to petroleum supply. The BSD goal is to develop technologies that are competitive with fossil fuels, in both cost and environmental performance, by the end of the decade. This document contains summaries of ongoing research sponsored by the DOE BSD. A summary sheet is presented for each project funded or in existence during FY 1993. Each summary sheet contains and account of project funding, objectives, accomplishments and current status, and significant publications.

  13. Computerized summary scoring: crowdsourcing-based latent semantic analysis.

    Li, Haiying; Cai, Zhiqiang; Graesser, Arthur C

    2017-11-03

    In this study we developed and evaluated a crowdsourcing-based latent semantic analysis (LSA) approach to computerized summary scoring (CSS). LSA is a frequently used mathematical component in CSS, where LSA similarity represents the extent to which the to-be-graded target summary is similar to a model summary or a set of exemplar summaries. Researchers have proposed different formulations of the model summary in previous studies, such as pregraded summaries, expert-generated summaries, or source texts. The former two methods, however, require substantial human time, effort, and costs in order to either grade or generate summaries. Using source texts does not require human effort, but it also does not predict human summary scores well. With human summary scores as the gold standard, in this study we evaluated the crowdsourcing LSA method by comparing it with seven other LSA methods that used sets of summaries from different sources (either experts or crowdsourced) of differing quality, along with source texts. Results showed that crowdsourcing LSA predicted human summary scores as well as expert-good and crowdsourcing-good summaries, and better than the other methods. A series of analyses with different numbers of crowdsourcing summaries demonstrated that the number (from 10 to 100) did not significantly affect performance. These findings imply that crowdsourcing LSA is a promising approach to CSS, because it saves human effort in generating the model summary while still yielding comparable performance. This approach to small-scale CSS provides a practical solution for instructors in courses, and also advances research on automated assessments in which student responses are expected to semantically converge on subject matter content.

  14. Worldwide Airfield Summary

    National Oceanic and Atmospheric Administration, Department of Commerce — The Worldwide Airfield Summary contains a selection of climatological data produced by the U.S. Air Force, Air Weather Service. The reports were compiled from dozens...

  15. Annual Meteorological Summaries

    National Oceanic and Atmospheric Administration, Department of Commerce — Single-year summaries of observations at Weather Bureau and cooperative stations across the United States. Predominantly the single page Form 1066, which includes...

  16. Summary of Research 1997

    Maier, William B.; Cleary, David D.

    1997-01-01

    This report contains summaries of research projects in the Department of Physics. A list of recent publications in also included which consists of conference presentations and publications, books, contributions to books, published jounal papers, technical reports, and thesis abstracts.

  17. Global Climate Summaries

    National Oceanic and Atmospheric Administration, Department of Commerce — The Global Hourly Summaries are simple indicators of observational normals which include climatic data summarizations and frequency distributions. These typically...

  18. Cancer Information Summaries

    Peer-reviewed, evidence-based summaries on topics including adult and pediatric cancer treatment, supportive and palliative care, screening, prevention, genetics, and complementary and alternative medicine. References to published literature are included.

  19. Summaries of poster contributions

    1981-01-01

    The 10. meeting covered subjects on the application of electron microscopy in numerous fields such as biology and medicine, solid state physics, semiconductor research and production, crystallography, materials science, and chemistry of polymers. 174 summaries of poster contributions are included

  20. Oceanographic Monthly Summary

    National Oceanic and Atmospheric Administration, Department of Commerce — Oceanographic Monthly Summary contains sea surface temperature (SST) analyses on both regional and ocean basin scales for the Atlantic, Pacific, and Indian Oceans....

  1. MSIS State Summary Datamarts

    U.S. Department of Health & Human Services — This page provides background needed to take advantage of the capabilities of the MSIS State Summary Datamart. This mart allows the user to develop high-level...

  2. Summary of blog

    Reader, Capitol

    2013-01-01

    This ebook consists of a summary of the ideas, viewpoints and facts presented by Hugh Hewitt in his book "Blog: Understanding the Information Reformation that's Changing Your World". This summary offers a concise overview of the entire book in less than 30 minutes reading time. However this work does not replace in any case Hugh Hewitt's book.Hewitt argues that blogs have an important potential and he believes that it would be a dreadful mistake to avoid their power.

  3. Site environmental report summary

    1992-01-01

    In this summary of the Fernald 1992 Site Environmental Report the authors will describe the impact of the Fernald site on man and the environment and provide results from the ongoing Environmental Monitoring Program. Also included is a summary of the data obtained from sampling conducted to determine if the site complies with DOE, US Environmental Protection Agency (USEPA), and Ohio EPA (OEPA) requirements. These requirements are set to protect both man and the environment

  4. An Attack Model Development Process for the Cyber Security of Safety Related Nuclear Digital I and C Systems

    Khand, Parvaiz Ahmed; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2007-10-15

    Nuclear power plants (NPPs), the redundant safety related systems are designed to take automatic action to prevent and mitigate accident conditions if the operators and the non-safety systems fail to maintain the plant within normal operating conditions. Presently, there is trend of connecting computer networks of commercial NPPs to corporate local area networks (LANs) to give engineers access to plant data for economic benefits. An increase in plant efficiency of a couple percentage points can translate to millions upon millions of dollars per year. The nuclear industry is also moving in the direction of installing digital controls that would allow for remote operation of plant functions, perhaps within a few years. However, this connectivity may also cause new security problems such as: in 2003, a computer worm named as slammer penetrated a private computer network at Ohio's Davis-Besse nuclear plant and disabled a safety monitoring system called a safety parameter display system (SPDS). Moreover, the present systems were developed with consideration of reliability and safety rather than security. In present scenario, there is a need to model and understand the cyber attacks towards these systems in a systematic way, and to demonstrate that the plant specific procedures and the imposed security controls adequately protect the systems from analyzed cyber security attacks. Attack trees provide a systematic, disciplined and effective way to model and understand cyber attacks towards any type of systems, make it possible to understand risks from deliberate, malicious intrusions from attackers, and make security decisions. Using attack trees the security of large systems can be modeled by considering a security breach as a system failure, and describing it with a set of events that can lead to system failure in a combinatorial way. The attacks towards the system are represented in a tree structure, with an attack that can significantly damage the system operation

  5. An Attack Model Development Process for the Cyber Security of Safety Related Nuclear Digital I and C Systems

    Khand, Parvaiz Ahmed; Seong, Poong Hyun

    2007-01-01

    Nuclear power plants (NPPs), the redundant safety related systems are designed to take automatic action to prevent and mitigate accident conditions if the operators and the non-safety systems fail to maintain the plant within normal operating conditions. Presently, there is trend of connecting computer networks of commercial NPPs to corporate local area networks (LANs) to give engineers access to plant data for economic benefits. An increase in plant efficiency of a couple percentage points can translate to millions upon millions of dollars per year. The nuclear industry is also moving in the direction of installing digital controls that would allow for remote operation of plant functions, perhaps within a few years. However, this connectivity may also cause new security problems such as: in 2003, a computer worm named as slammer penetrated a private computer network at Ohio's Davis-Besse nuclear plant and disabled a safety monitoring system called a safety parameter display system (SPDS). Moreover, the present systems were developed with consideration of reliability and safety rather than security. In present scenario, there is a need to model and understand the cyber attacks towards these systems in a systematic way, and to demonstrate that the plant specific procedures and the imposed security controls adequately protect the systems from analyzed cyber security attacks. Attack trees provide a systematic, disciplined and effective way to model and understand cyber attacks towards any type of systems, make it possible to understand risks from deliberate, malicious intrusions from attackers, and make security decisions. Using attack trees the security of large systems can be modeled by considering a security breach as a system failure, and describing it with a set of events that can lead to system failure in a combinatorial way. The attacks towards the system are represented in a tree structure, with an attack that can significantly damage the system operation as a

  6. Do you see what I see? Effects of national culture on employees' safety-related perceptions and behavior.

    Casey, Tristan W; Riseborough, Karli M; Krauss, Autumn D

    2015-05-01

    Growing international trade and globalization are increasing the cultural diversity of the modern workforce, which often results in migrants working under the management of foreign leadership. This change in work arrangements has important implications for occupational health and safety, as migrant workers have been found to be at an increased risk of injuries compared to their domestic counterparts. While some explanations for this discrepancy have been proposed (e.g., job differences, safety knowledge, and communication difficulties), differences in injury involvement have been found to persist even when these contextual factors are controlled for. We argue that employees' national culture may explain further variance in their safety-related perceptions and safety compliance, and investigate this through comparing the survey responses of 562 Anglo and Southern Asian workers at a multinational oil and gas company. Using structural equation modeling, we firstly established partial measurement invariance of our measures across cultural groups. Estimation of the combined sample structural model revealed that supervisor production pressure was negatively related to willingness to report errors and supervisor support, but did not predict safety compliance behavior. Supervisor safety support was positively related to both willingness to report errors and safety compliance. Next, we uncovered evidence of cultural differences in the relationships between supervisor production pressure, supervisor safety support, and willingness to report errors; of note, among Southern Asian employees the negative relationship between supervisor production pressure and willingness to report errors was stronger, and for supervisor safety support, weaker as compared to the model estimated with Anglo employees. Implications of these findings for safety management in multicultural teams within the oil and gas industry are discussed. Copyright © 2015 Elsevier Ltd. All rights reserved.

  7. Green fluorescent protein labeling of Listeria, Salmonella, and Escherichia coli O157:H7 for safety-related studies.

    Li Ma

    Full Text Available Many food safety-related studies require tracking of introduced foodborne pathogens to monitor their fate in complex environments. The green fluorescent protein (GFP gene (gfp provides an easily detectable phenotype so has been used to label many microorganisms for ecological studies. The objectives of this study were to label major foodborne pathogens and related bacteria, including Listeria monocytogenes, Listeria innocua, Salmonella, and Escherichia coli O157:H7 strains, with GFP and characterize the labeled strains for stability of the GFP plasmid and the plasmid's effect on bacterial growth. GFP plasmids were introduced into these strains by a CaCl(2 procedure, conjugation or electroporation. Stability of the label was determined through sequential propagation of labeled strains in the absence of selective pressure, and rates of plasmid-loss were calculated. Stability of the GFP plasmid varied among the labeled species and strains, with the most stable GFP label observed in E. coli O157:H7. When grown in nonselective media for two consecutive subcultures (ca. 20 generations, the rates of plasmid loss among labeled E. coli O157:H7, Salmonella and Listeria strains ranged from 0%-30%, 15.8%-99.9% and 8.1%-93.4%, respectively. Complete loss (>99.99% of the plasmid occurred in some labeled strains after five consecutive subcultures in the absence of selective pressure, whereas it remained stable in others. The GFP plasmid had an insignificant effect on growth of most labeled strains. E. coli O157:H7, Salmonella and Listeria strains can be effectively labeled with the GFP plasmid which can be stable in some isolates for many generations without adversely affecting growth rates.

  8. ITER plasma facing components

    Kuroda, T.; Vieider, G.; Akiba, M.

    1991-01-01

    This document summarizes results of the Conceptual Design Activities (1988-1990) for the International Thermonuclear Experimental Reactor (ITER) project, namely those that pertain to the plasma facing components of the reactor vessel, of which the main components are the first wall and the divertor plates. After an introduction and an executive summary, the principal functions of the plasma-facing components are delineated, i.e., (i) define the low-impurity region within which the plasma is produced, (ii) absorb the electromagnetic radiation and charged-particle flux from the plasma, and (iii) protect the blanket/shield components from the plasma. A list of critical design issues for the divertor plates and the first wall is given, followed by discussions of the divertor plate design (including the issues of material selection, erosion lifetime, design concepts, thermal and mechanical analysis, operating limits and overall lifetime, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, and advanced divertor concepts) and the first wall design (armor material and design, erosion lifetime, overall design concepts, thermal and mechanical analysis, lifetime and operating limits, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, an alternative first wall design, and the limiters used instead of the divertor plates during start-up). Refs, figs and tabs

  9. Overview and summary

    1999-01-01

    The ITER Physics Basis presents and evaluates the physics rules and methodologies for plasma performance projections, which provide the basis for the design of a tokamak burning plasma device whose goal is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. This Chapter summarizes the physics basis for burning plasma projections, which is developed in detail by the ITER Physics Expert Groups in subsequent chapters. To set context, the design guidelines and requirements established in the report of ITER Special Working Group 1 are presented, as are the specifics of the tokamak design developed in the Final Design Report of the ITER Engineering Design Activities, which exemplifies burning tokamak plasma experiments. The behaviour of a tokamak plasma is determined by the interaction of many diverse physics processes, all of which bear on projections for both a burning plasma experiment and an eventual tokamak reactor. Key processes summarized here are energy and particle confinement and the H-mode power threshold; MHD stability, including pressure and density limits, neoclassical islands, error fields, disruptions, sawteeth, and ELMs; power and particle exhaust, involving divertor power dispersal, helium exhaust, fuelling and density control, H-mode edge transition region, erosion of plasma facing components, tritium retention; energetic particle physics; auxiliary power physics; and the physics of plasma diagnostics. Summaries of projection methodologies, together with estimates of their attendant uncertainties, are presented in each of these areas. Since each physics element has its own scaling properties, an integrated experimental demonstration of the balance between the combined processes which obtains in a reactor plasma is inaccessible to contemporary experimental facilities: it requires a reactor scale device. It is argued, moreover, that a burning plasma experiment can be sufficiently flexible to permit

  10. Microgrid Design Toolkit (MDT) Technical Documentation and Component Summaries

    Arguello, Bryan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gearhart, Jared Lee [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jones, Katherine A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Eddy, John P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The Microgrid Design Toolkit (MDT) is a decision support software tool for microgrid designers to use during the microgrid design process. The models that support the two main capabilities in MDT are described. The first capability, the Microgrid Sizing Capability (MSC), is used to determine the size and composition of a new microgrid in the early stages of the design process. MSC is a mixed-integer linear program that is focused on developing a microgrid that is economically viable when connected to the grid. The second capability is focused on refining a microgrid design for operation in islanded mode. This second capability relies on two models: the Technology Management Optimization (TMO) model and Performance Reliability Model (PRM). TMO uses a genetic algorithm to create and refine a collection of candidate microgrid designs. It uses PRM, a simulation based reliability model, to assess the performance of these designs. TMO produces a collection of microgrid designs that perform well with respect to one or more performance metrics.

  11. Concept of a new method for fatigue monitoring of nuclear power plant components

    Zafosnik, M.; Cizelj, L.

    2007-01-01

    Fatigue is one of the well-understood aging mechanisms affecting mechanical components in many industrial facilities including nuclear power plants. Operational experience of nuclear power plants worldwide to date confirmed adequate design of safety related components against fatigue. In some cases however, for example when the plant life extension is envisioned, it may be very useful to monitor the remaining fatigue life of safety related components. Nuclear power plants components are classified into safety classes regarding their importance in mitigating the consequences of hypothetic accidents. Service life of components subjected to fatigue loading can be estimated with Usage Factor uk. A concept of the new method aiming both at monitoring the current state of the component and predicting its remaining lifetime in the life-extension conditions is presented. The method is based on determination of partial Usage Factor of components in which operating transients will be considered and compared to design transients. (author)

  12. Reliability Prediction Of System And Component Of Process System Of RSG-GAS Reactor

    Sitorus Pane, Jupiter

    2001-01-01

    The older the reactor the higher the probability of the system and components suffer from loss of function or degradation. This phenomenon occurred because of wear, corrosion, and fatigue. Study on component reliability was generally performed deterministically and statistically. This paper would describe an analysis of using statistical method, i.e. regression Cox, in order to predict the reliability of the components and their environmental influence's factors. The result showed that the dynamics, non safety related, and mechanic components have higher risk of failure, whereas static, safety related, and electric have lower risk of failures. The relative risk value for variable of components dynamics, quality, dummy 1 and dummy 2 are of 1.54, 1.59, 1.50, and 0.83 compare to other components type with each variable. Component with the higher risk have lower reliability than lower one

  13. Principal components

    Hallin, M.; Hörmann, S.; Piegorsch, W.; El Shaarawi, A.

    2012-01-01

    Principal Components are probably the best known and most widely used of all multivariate analysis techniques. The essential idea consists in performing a linear transformation of the observed k-dimensional variables in such a way that the new variables are vectors of k mutually orthogonal

  14. Indexing for summary queries

    Yi, Ke; Wang, Lu; Wei, Zhewei

    2014-01-01

    ), of a particular attribute of these records. Aggregation queries are especially useful in business intelligence and data analysis applications where users are interested not in the actual records, but some statistics of them. They can also be executed much more efficiently than reporting queries, by embedding...... returned by reporting queries. In this article, we design indexing techniques that allow for extracting a statistical summary of all the records in the query. The summaries we support include frequent items, quantiles, and various sketches, all of which are of central importance in massive data analysis....... Our indexes require linear space and extract a summary with the optimal or near-optimal query cost. We illustrate the efficiency and usefulness of our designs through extensive experiments and a system demonstration....

  15. MIV Project: Executive Summary

    Ravazzotti, Mariolina T.; Jørgensen, John Leif; Neefs, Marc

    1997-01-01

    Under the ESA contract #11453/95/NL/JG(SC), aiming at assessing the feasibility of Rendez-vous and docking of unmanned spacecrafts, a reference mission scenario was defined. This report gives an executive summary of the achievements and results from the project.......Under the ESA contract #11453/95/NL/JG(SC), aiming at assessing the feasibility of Rendez-vous and docking of unmanned spacecrafts, a reference mission scenario was defined. This report gives an executive summary of the achievements and results from the project....

  16. Crisis Management: Research Summaries

    Brock, Stephen E., Ed.; Dorman, Sally; Anderson, Luke; McNair, Daniel

    2013-01-01

    This article presents summaries of three studies relevant to school crisis response. The first report, "A Framework for International Crisis Intervention" (Sally Dorman), is a review of how existing crisis intervention models (including the NASP PREPaRE model) have been adapted for international use. The second article, "Responding…

  17. Summary of Trends

    First page Back Continue Last page Overview Graphics. Summary of Trends. Optical Ethernet: Direct Ethernet connectivity to businesses through optical fiber. Automation of network infrastructure: Cross-connects for interconnections; Intelligence through software for OA&M. New “data-centric” protection mechanisms ...

  18. Summary of discussion

    2006-01-01

    This document provides summaries of the discussions occurred during the second international workshop on the indemnification of nuclear damage. It concerns the second accident scenario: a fire on board of a ship transporting enriched uranium hexafluoride along the Danube River. (A.L.B.)

  19. Geothermal energy. Program summary

    1979-06-01

    Brief descriptions of geothermal projects funded through the Department of Energy during FY 1978 are presented. Each summary gives the project title, contractor name, contract number, funding level, dates, location, and name of the principal investigator, together with project highlights, which provide informaion such as objectives, strategies, and a brief project description. (MHR)

  20. Symposium summary and prognosis

    Bjorken, J.D.

    1975-11-01

    The summary of the symposium on high energy physics experiments includes phenomena at low energies, the foundations of physics (considered to be mainly gravitation and quantum electrodynamics), standards of reference used for interpretation of experimental data, the new physics, particle proliferation, theoretical development, and a prognosis for the future

  1. Summary of Session III

    Furman, M.A.

    2002-01-01

    This is a summary of the talks presented in Session III ''Simulations of Electron-Cloud Build Up'' of the Mini-Workshop on Electron-Cloud Simulations for Proton and Positron Beams ECLOUD-02, held at CERN, 15-18 April 2002

  2. Summaries and future projections

    Egelstaff, P.A.

    1989-01-01

    In this paper the author gives a brief summary of this meeting. He discusses the status at the current neutron sources and future sources. The current problems with targets, moderators, performance of storage rings and shields are briefly mentioned. Finally, he speculates on the prospects of neutron sources for the future and gives his version of the ultimate source

  3. Executive Summaries: CIL '90.

    Elsweiler, John A., Jr.; And Others

    1990-01-01

    Presents summaries of 12 papers presented at the 1990 Computers in Libraries Conference. Topics discussed include online searching; microcomputer-based serials management; microcomputer-based workstations; online public access catalogs (OPACs); multitype library networking; CD-ROM searches; locally mounted online databases; collection evaluation;…

  4. Healthcare. Executive Summary

    Carnevale, Anthony P.; Smith, Nicole; Gulish, Artem; Beach, Bennett H.

    2012-01-01

    This executive summary highlights several findings about healthcare. These are: (1) Healthcare is 18 percent of the U.S. economy, twice as high as in other countries; (2) There are two labor markets in healthcare: high-skill, high-wage professional and technical jobs and low-skill, low-wage support jobs; (3) Demand for postsecondary education in…

  5. Summary Stage 2018 - SEER

    Access this manual of codes and coding instructions for the summary stage field for cases diagnosed January 1, 2018 and forward. 2018 version applies to every site and/or histology combination, including lymphomas and leukemias. Historically, also called General Staging, California Staging, and SEER Staging.

  6. Work stress and patient safety: observer-rated work stressors as predictors of characteristics of safety-related events reported by young nurses.

    Elfering, A; Semmer, N K; Grebner, S

    This study investigates the link between workplace stress and the 'non-singularity' of patient safety-related incidents in the hospital setting. Over a period of 2 working weeks 23 young nurses from 19 hospitals in Switzerland documented 314 daily stressful events using a self-observation method (pocket diaries); 62 events were related to patient safety. Familiarity of safety-related events and probability of recurrence, as indicators of non-singularity, were the dependent variables in multilevel regression analyses. Predictor variables were both situational (self-reported situational control, safety compliance) and chronic variables (job stressors such as time pressure, or concentration demands and job control). Chronic work characteristics were rated by trained observers. The most frequent safety-related stressful events included incomplete or incorrect documentation (40.3%), medication errors (near misses 21%), delays in delivery of patient care (9.7%), and violent patients (9.7%). Familiarity of events and probability of recurrence were significantly predicted by chronic job stressors and low job control in multilevel regression analyses. Job stressors and low job control were shown to be risk factors for patient safety. The results suggest that job redesign to enhance job control and decrease job stressors may be an important intervention to increase patient safety.

  7. Literature review of environmental qualification of safety-related electric cables: Literature analysis and appendices. Volume 2

    Lofaro, R.; Bowerman, B.; Carbonaro, J.

    1996-04-01

    In support of the US NRC Environmental Qualification (EQ) Research Program, a literature review was performed to identify past relevant work that could be used to help fully or partially resolve issues of interest related to the qualification of low-voltage electric cable. A summary of the literature reviewed is documented in Volume 1 of this report. In this, Volume 2 of the report, dossiers are presented which document the issues selected for investigation in this program, along with recommendations for future work to resolve the issues, when necessary. The dossiers are based on an analysis of the literature reviewed, as well as expert opinions. This analysis includes a critical review of the information available from past and ongoing work in thirteen specific areas related to EQ. The analysis for each area focuses on one or more questions which must be answered to consider a particular issue resolved. Results of the analysis are presented, along with recommendations for future work. The analysis is documented in the form of a dossier for each of the areas analyzed

  8. Inelastic seismic behavior of post-installed anchors for nuclear safety related structures: Generation of experimental database

    Mahadik, Vinay, E-mail: vinay.mahadik@iwb.uni-stuttgart.de; Sharma, Akanshu; Hofmann, Jan

    2016-02-15

    Highlights: • Experiments for evaluating seismic behavior of anchors were performed. • Two undercut anchor products in use in nuclear facilities were considered. • Monotonic tension, shear and cycling tension tests at different crack widths. • Crack cycling tests at constant, in-phase and out-of phase tension loads. • Characteristics for the two anchors as a function of crack width were identified. - Abstract: Post installed (PI) anchors are often employed for connections between concrete structure and components or systems in nuclear power plants (NPP) and related facilities. Standardized practices for nuclear related structures demand stringent criteria, which an anchor has to satisfy in order to qualify for use in NPP related structures. In NPP and related facilities, the structure–component interaction in the event of an earthquake depends on the inelastic behavior of the concrete structure, the component system and also the anchorage system that connects them. For analysis, anchorages are usually assumed to be rigid. Under seismic actions, however, it is known that anchors may undergo significant plastic displacement and strength degradation. Analysis of structure–component interaction under seismic loads calls for numerical models simulating inelastic behavior of anchorage systems. A testing program covering different seismic loading scenarios in a reasonably conservative manner is required to establish a basis for generating numerical models of anchorage systems. Currently there is a general lack of modeling techniques to consider the inelastic behavior of anchorages in structure–component interaction under seismic loads. In this work, in view of establishing a basis for development of numerical models simulating the inelastic behavior of anchors, seismic tests on two different undercut anchors qualified for their use in NPP related structures were carried out. The test program was primarily based on the DIBt-KKW-Leitfaden (2010) guidelines

  9. Mirror Confinement Systems: project summaries

    1980-07-01

    This report contains descriptions of the projects supported by the Mirror Confinement Systems (MCS) Division of the Office of Fusion Energy. The individual project summaries were prepared by the principal investigators, in collaboration with MCS staff office, and include objectives and milestones for each project. In addition to project summaries, statements of Division objectives and budget summaries are also provided

  10. Summary on experiments

    Livingston, A.E.

    1981-01-01

    Experimental studies of the atomic structures of both simple and complex atoms and ions provide crucial tests of atomic structure theory and of calculational techniques for a wide range of atomic systems. This summary is restricted to a brief discussion of some recent and current experiments in few-electron and many-electron atoms and ions which represent exciting challenges to sophisticated atomic structure calculations, discussed elsewhere. In particular the emphasis is on high-Z systems

  11. Blois V: Experimental summary

    Albrow, M.G.

    1993-09-01

    The author gives a summary talk of the best experimental data given at the Vth Blois Workshop on Elastic and Diffractive Scattering. He addresses the following eight areas in his talk: total and elastic cross sections; single diffractive excitation; electron-proton scattering; di-jets and rapidity gaps; areas of future study; spins and asymmetries; high-transverse momentum and masses at the Tevatron; and disoriented chiral condensates and cosmic radiation

  12. Summary and outlook

    Jong, M. de, E-mail: mjg@nikhef.nl [Nikhef - National Institute for Subatomic Physics, Science Park 105, 1098 XG Amsterdam (Netherlands); LION - Leiden Institute of Physics, Leiden University, PO Box 9504, 2300 RA Leiden (Netherlands)

    2013-10-11

    In 2003, a series of Very Large Volume Neutrino Telescope Workshops (VLVnT) was initiated in Amsterdam, the Netherlands. The 5th workshop in this series took place in Erlangen, Germany, between 12–14 October 2011 and focused on the aspects of high-energy neutrino astronomy. In this summary report, an overview of the activities world-wide is presented as well as the perspectives of the field.

  13. Blois 5: Experimental summary

    Albrow, M. G.

    1993-09-01

    The author gives a summary talk of the best experimental data given at the 5th Blois Workshop on Elastic and Diffractive Scattering. He addresses the following eight areas in his talk: total and elastic cross sections; single diffractive excitation; electron-proton scattering; di-jets and rapidity gaps; areas of future study; spins and asymmetries; high-transverse momentum and masses at the Tevatron; and disoriented chiral condensates and cosmic radiation.

  14. Mineral commodity summaries 2015

    ,

    2015-01-01

    Each chapter of the 2015 edition of the U.S. Geological Survey (USGS) Mineral Commodity Summaries (MCS) includes information on events, trends, and issues for each mineral commodity as well as discussions and tabular presentations on domestic industry structure, Government programs, tariffs, 5-year salient statistics, and world production and resources. The MCS is the earliest comprehensive source of 2014 mineral production data for the world. More than 90 individual minerals and materials are covered by two-page synopses.

  15. Blois V: Experimental summary

    Albrow, M.G.

    1993-09-01

    The author gives a summary talk of the best experimental data given at the Vth Blois Workshop on Elastic and Diffractive Scattering. He addresses the following eight areas in his talk: total and elastic cross sections; single diffractive excitation; electron-proton scattering; di-jets and rapidity gaps; areas of future study; spins and asymmetries; high-transverse momentum and masses at the Tevatron; and disoriented chiral condensates and cosmic radiation.

  16. ULSGEN (Uplink Summary Generator)

    Wang, Y.-F.; Schrock, M.; Reeve, T.; Nguyen, K.; Smith, B.

    2014-01-01

    Uplink is an important part of spacecraft operations. Ensuring the accuracy of uplink content is essential to mission success. Before commands are radiated to the spacecraft, the command and sequence must be reviewed and verified by various teams. In most cases, this process requires collecting the command data, reviewing the data during a command conference meeting, and providing physical signatures by designated members of various teams to signify approval of the data. If commands or sequences are disapproved for some reason, the whole process must be restarted. Recording data and decision history is important for traceability reasons. Given that many steps and people are involved in this process, an easily accessible software tool for managing the process is vital to reducing human error which could result in uplinking incorrect data to the spacecraft. An uplink summary generator called ULSGEN was developed to assist this uplink content approval process. ULSGEN generates a web-based summary of uplink file content and provides an online review process. Spacecraft operations personnel view this summary as a final check before actual radiation of the uplink data. .

  17. Nuclear plant reliability data system. 1979 annual reports of cumulative system and component reliability

    1979-01-01

    The primary purposes of the information in these reports are the following: to provide operating statistics of safety-related systems within a unit which may be used to compare and evaluate reliability performance and to provide failure mode and failure rate statistics on components which may be used in failure mode effects analysis, fault hazard analysis, probabilistic reliability analysis, and so forth

  18. The effectiveness of a bicycle safety program for improving safety-related knowledge and behavior in young elementary students.

    McLaughlin, Karen A; Glang, Ann

    2010-05-01

    The purpose of this study was to evaluate the "Bike Smart" program, an eHealth software program that teaches bicycle safety behaviors to young children. Participants were 206 elementary students in grades kindergarten to 3. A random control design was employed to evaluate the program, with students assigned to either the treatment condition (Bike Smart) or the control condition (a video on childhood safety). Outcome measures included computer-based knowledge items (safety rules, helmet placement, hazard discrimination) and a behavioral measure of helmet placement. Results demonstrated that regardless of gender, cohort, and grade the participants in the treatment group showed greater gains than control participants in both the computer-presented knowledge items (p > .01) and the observational helmet measure (p > .05). Findings suggest that the Bike Smart program can be a low cost, effective component of safety training packages that include both skills-based and experiential training.

  19. LEAP 1992: Conference summary

    Dover, C.B.

    1992-12-01

    We present a summary of the many new results in antiproton (bar p) physics presented at the LEAP '92 conference, in the areas of meson spectroscopy, bar NN scattering, annihilation and spin observables, strangeness and charm production, bar N annihilation in nuclei, atomic physics with very low energy bar p's, the exploration of fundamental symmetries and interactions with bar p (CP, T, CPT, gravitation), and the prospects for new bar p facilities at ultralow energies or energies above the LEAR regime (≥ 2 GeV/c)

  20. FY 1996 activity summary

    1997-01-01

    The US Department of Energy Office of Nuclear and Facility Safety provides nuclear safety policy, independent technical evaluation, and technical support. A summary of these activities is provided in this report. These include: (1) changing the mission of the former production facilities to storage and waste management; (2) stabilizing nuclear materials not recycled due to production cessation or interruptions; (3) reformulating the authorization basis for existing facilities to convert to a standards based approach for operations consistent with modern expectations; and (4) implementing a modern regulatory framework for nuclear facilities. Enforcement of the Price-Anderson Amendments Act is also reported

  1. Program summary report

    1978-01-01

    The report provides summary information on all phases of nuclear regulation, and is intended as an information and decision-making tool for mid and upper level management of the Nuclear Regulatory Commission. The report is divided functionally into ten sections: (1) nuclear power plants in the United States; (2) operating nuclear power plants; (3) reactors under construction; (4) operating license applications under NRC review; (5) construction permit applications and special projects under NRC review; (6) ACRS and ASLBP; (7) nuclear materials; (8) standards and regulations; (9) research projects; and (10) foreign reactors

  2. Summary of main points

    2006-01-01

    In conjunction with its 6. annual meeting, the WPDD in close co-operation with the FSC held a Topical session on 'Stakeholder Involvement in Decommissioning' on November 14, 2005. The session was attended by 36 participants totally representing 14 NEA member countries and 2 international organisations. Two keynote addresses were given at the Topical Session. The first one treated of what is needed for robust decisions and how to bring all stakeholders into the debate. In the second keynote address a summary was made on what have been said on stakeholder involvement in decommissioning during earlier meetings of the WPDD. The main part of the session was then devoted to views from different stakeholders regarding their role and their involvement. This part contained viewpoints from local communities (Kaevlinge in Sweden and Port Hope in Canada), authorities (Scottish Executive and CSNC) and operators (EDF from France and EWN from Germany). Case studies from the decommissioning of Dounrey in the UK and from Trojan and Main Yankee in the USA were presented in the end part of the Topical session followed by a summary and lessons learnt report by the Rapporteur. A detailed programme of the Topical session can be seen in Appendix 1

  3. Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

    Boardman, J.R.

    1994-10-01

    This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures

  4. Thermal overload protection for electric motors on safety-related motor-operated valves: Generic Issue II.E.6.1

    Rothberg, O.

    1988-06-01

    NRC regulatory positions, as stated in Regulatory Guide 1.106, Revision 1, have been identified by the Office for Analysis and Evaluation of Operational Data (AEOD) as potential contributors to valve motor burnout. AEOD is particularly concerned about the allowed policy of bypassing thermal overload devices during normal or accident conditions. Regulatory Guide 1.106 favors compromising the function of thermal overload devices in favor of completing the safety-related action of valves. The purpose of this study was to determine if the guidance contained in Regulatory Guide 1.106 is appropriate and, if not, to recommend the necessary changes. This report describes thermal overload devices commonly used to protect safety-related valve operator motors. The regulatory guidelines stated in Regulatory Guide 1.106 along with the limitations of thermal overload protection are discussed. Supplements and alternatives to thermal overload protection are also described. Findings and conclusions of several AEOD reports are discussed. Information obtained from the standard review plan, standard technical specifications, technical specifications from representative plants, and several papers are cited

  5. Hazmat Yearly Incident Summary Reports

    Department of Transportation — Series of Incident data and summary statistics reports produced which provide statistical information on incidents by type, year, geographical location, and others....

  6. Neutrino physics: Summary talk

    Marciano, W.J.

    1989-04-01

    This paper is organized as follows: First, I describe the state of neutrino phenomenology. Emphasis is placed on sin 2 θ W , its present status and future prospects. In addition, some signatures of ''new physics'' are described. Then, kaon physics at Fermilab is briefly discussed. I concentrate on the interesting rare decay K L → π 0 e + e - which may be a clean probe direct CP violation. Neutrino mass, mixing, and electromagnetic moments are surveyed. There, I describe the present state and future direction of accelerator based experiments. Finally, I conclude with an outlook on the future. Throughout this summary, I have drawn from and incorporated ideas discussed by other speakers at this workshop. However, I have tried to combine their ideas with my own perspective on neutrino physics and where it is headed. 49 refs., 3 figs., 4 tabs

  7. Summary of group discussions

    2009-01-01

    A key aspect of the workshop was the interaction and exchange of ideas and information among the 40 participants. To facilitate this activity the workshop participants were divided into five discussions groups. These groups reviewed selected subjects and reported back to the main body with summaries of their considerations. Over the 3 days the 5 discussion groups were requested to focus on the following subjects: the characteristics and capabilities of 'good' organisations; how to ensure sufficient resources; how to ensure competence within the organisation; how to demonstrate organisational suitability; the regulatory oversight processes - including their strengths and weaknesses. A list of the related questions that were provided to the discussion groups can be found in Appendix 3. Also included in Appendix 3 are copies of the slides the groups prepared that summarised their considerations

  8. Vacuum considerations: summary

    Blechschmidt, D.; Halama, H.J.

    1978-01-01

    A summary is given of the efforts of a vacuum systems study group of the workshop on a Heavy Ion Demonstration Experiment (HIDE) for heavy ion fusion. An inadequate knowledge of cross-sections prevents a more concrete vacuum system design. Experiments leading to trustworthy numbers for charge exchange, stripping and capture cross-sections are badly needed and should start as soon as possible. In linacs, beam loss will be almost directly proportional to the pressure inside the tanks. The tanks should, therefore, be built in such a way that they can be baked-out in situ to improve their vacuum, especially if the cross-sections turn out to be higher than anticipated. Using standard UHV techniques and existing pumps, an even lower pressure can be achieved. The vacuum system design for circular machines will be very difficult, and in some cases, beyond the present state-of-the-art

  9. Summary and conclusions

    2000-01-01

    The international workshop on 'Nuclear Power Plant Life Management in a Changing Business World' was held in Washington, DC, on 26-27 June 2000. This workshop was attended by more than 50 experts from 12 countries and three international organisations. The workshop included a series of presentations to a plenary session of all participants. A spectrum of experiences in plant life extension activities as well as experiences in operating a nuclear power plant (NPP) in a 'free-market' electricity environment were presented: The workshop also included three working groups in which major issues facing PLIM activities for NPPs were identified and discussed. The three working groups covered technology, regulation and business. The following sections of this report consist of summaries of the discussions that took place in each of the three working groups. (author)

  10. Summary: Hadron dynamics sessions

    Carroll, A.S.; Londergan, J.T.

    1993-01-01

    Four sessions on Hadron Dynamics were organized at this Workshop. The first topic, QCD Exclusive Reactions and Color Transparency, featured talks by Ralston, Heppelman and Strikman; the second, QCD and Inclusive Reactions had talks by Garvey, Speth and Kisslinger. The third dynamics session, Medium Modification of Elementary Interactions had contributions from Kopeliovich, Alves and Gyulassy; the fourth session Pre-QCD Dynamics and Scattering, had talks by Harris, Myhrer and Brown. An additional joint Spectroscopy/Dynamics session featured talks by Zumbro, Johnson and McClelland. These contributions are reviewed briefly in this summary. Two additional joint sessions between Dynamics and η physics are reviewed by the organizers of the Eta sessions. In such a brief review there is no way the authors can adequately summarize the details of the physics presented here. As a result, they concentrate only on brief impressionistic sketches of the physics topics discussed and their interrelations. They include no bibliography in this summary, but simply refer to the talks given in more detail in the Workshop proceedings. They focus on topics which were common to several presentations in these sessions. First, nuclear and particle descriptions of phenomena are now clearly converging, in both a qualitative and quantitative sense; they show several examples of this convergence. Second, an important issue in hadron dynamics is the extent to which elementary interactions are modified in nuclei at high energies and/or densities, and they illustrate some of these medium effects. Finally, they focus on those dynamical issues where hadron facilities can make an important, or even a unique, contribution to the knowledge of particle and nuclear physics

  11. Quantitative and qualitative analysis of student textbook summary writing

    Demaree, Dedra; Allie, Saalih; Low, Michael; Taylor, Julian

    2008-10-01

    The majority of "special access" students at the University of Cape Town are second language English speakers for whom reading the physics textbook is daunting. As a strategy to encourage meaningful engagement with the text, students wrote textbook summaries due the day material was covered in class. The summaries were returned, and they could bring them or re-write them for use during their examinations. A framework was developed to analyze the summaries based on Waywood, defining three cognitive levels seen in mathematics journaling: recounting, summarizing, and dialoging. This framework was refined, expanded, and tested. Interviews with students were conducted for their views on summary writing and survey questions were included on their final exams. The study was carried out in the 2007 spring semester of the "Foundation Physics Course," a component of the special access program.

  12. Screening of external hazards for NPP with bank type reactor. Modeling of safety related systems and equipment for RBMK. Probabilistic assessment of NPP safety on aircraft impact. Progress report

    Kostarev, V.

    1999-01-01

    This progress report was produced within the frame of IAEA research project on screening the hazards for NPP with bank type reactor. It covers the following tasks; development of the model for the primary loop system of RBMK; developing the models for safety related equipment of RBMK; developing of models for safety related models of EGP-6 type reactor (Bilibinskaya Nuclear Co-generated heat and Power Plant); and probabilistic assessment of NPP safety on aircraft impact

  13. Interim summary report of the safety case 2009

    2010-03-01

    intrinsic properties of the main components of the repository and from the understanding of their evolution gained from extensive site- and concept-specific field, laboratory and modelling studies and from studies of natural and anthropogenic analogues. For any canisters that fail over this time window, the low radionuclide calculated release rates to the biosphere and resultant annual effective doses to humans and absorbed dose rates to other species of flora and fauna imply that any radiological consequences of these releases will be negligible. Furthermore, the calculation results indicate that, in general, differences in the geometry and transport paths considered in the analyses of the KBS-3V and KBS-3H design variants have only a minor impact on calculated releases and doses. Work carried out to date indicates that a geological repository for the final disposal of spent fuel, implemented as planned at the Olkiluoto site, will conform to Finnish regulatory requirements and provide an adequate level of longterm safety. This conclusion is based on the findings of safety assessments, the systematic treatment of uncertainty in these assessments and the quality measures that have been applied in the development and application of models, data and computer codes. Plans are in place to manage remaining safety-related issues and uncertainties, as given in the report TKS-2009. In implementing TKS-2009, quality assurance measures will be applied in the various production steps of the safety case, including and tests and experiments to demonstrate the feasibility and quality of technical solutions. In this way, a comprehensive safety case will be developed to support the licensing process. (orig.)

  14. "Against the silence": Development and first results of a patient survey to assess experiences of safety-related events in hospital

    Schwappach David LB

    2008-03-01

    Full Text Available Abstract Background Involvement of patients in the detection and prevention of safety related events and medical errors have been widely recommended. However, it has also been questioned whether patients at large are willing and able to identify safety-related events in their care. The aim of this study was to develop and pilot test a brief patient safety survey applicable to inpatient care in Swiss hospitals. Methods A survey instrument was developed in an iterative procedure. The instrument asks patients to report whether they have experienced specific undesirable events during their hospital stay. The preliminary version was developed together with experts and tested in focus groups with patients. The adapted survey instrument was pilot-tested in random samples of patients of two Swiss hospitals (n = 400. Responders to the survey that had reported experience of any incident were sampled for qualitative interviews (n = 18. Based on the interview, the researcher classified the reported incidents as confirmed or discarded. Results The survey was generally well accepted in the focus groups and interviews. In the quantitative pilot test, 125 patients returned the survey (response rate: 31%. The mean age of responders was 55 years (range 17–91, SD 18 years and 62.5% were female. The 125 participating patients reported 94 "definitive" and 34 "uncertain" events. 14% of the patients rated any of the experienced events as "serious". The definitive and uncertain events reported with highest frequency were phlebitis, missing hand hygiene, allergic drug reaction, unavailability of documents, and infection. 23% of patients reported some or serious concerns about their safety. The qualitative interviews indicate that both, the extent of patients' uncertainty in the classification of events and the likelihood of confirmation by the interviewer vary very much by type of incident. Unexpectedly, many patients reported problems and incidents related to food

  15. Impact of Safety-Related Regulations on Codeine Use in Children: A Quasi-Experimental Study Using Taiwan's National Health Insurance Research Database.

    Lin, Chih-Wan; Wang, Ching-Huan; Huang, Wei-I; Ke, Wei-Ming; Chao, Pi-Hui; Chen, Wen-Wen; Hsiao, Fei-Yuan

    2017-07-01

    Safety concerns regarding potential life-threatening adverse events associated with codeine have resulted in policy decisions to restrict its use in pediatrics. However, whether these drug safety communications have had an immediate and strong impact on codeine use remains in question. We aimed to investigate the impact of the two implemented safety-related regulations (label changes and reimbursement regulations) on the use of codeine for upper respiratory infection (URI) or cough. A quasi-experimental study was performed using Taiwan's National Health Insurance Research Database. Quarterly data of codeine prescription rates for URI/cough visits were reported, and an interrupted time series design was used to assess the impact of the safety regulations on the uses of codeine among children with URI/cough visits. Multivariable logistic regression models were used to explore patient and provider characteristics associated with the use of codeine. The safety-related regulations were associated with a significant reduction in codeine prescription rates of -4.24% (95% confidence interval [CI] -4.78 to -3.70), and the relative reduction compared with predicted rates based on preregulation projections was 60.4, 56.6, and 53.2% in the first, second, and third year after the regulations began, respectively. In the postregulation period, physicians specializing in otolaryngology (odds ratio [OR] 1.47, 95% CI 1.45-1.49), practicing in district hospitals (OR 6.84, 95% CI 5.82-8.04) or clinics (OR 6.50, 95% CI 5.54-7.62), and practicing in the least urbanized areas (OR 1.60, 95% CI 1.55-1.64) were more likely to prescribe codeine to children than their counterparts. Our study provides a successful example of how to effectively reduce the codeine prescriptions in children in the 'real-world' settings, and highlights areas where future effort could be made to improve the safety use of codeine. Future research is warranted to explore whether there was a simultaneous decrease in

  16. Component Cooling Heat Exchanger Heat Transfer Capability Operability Monitoring

    Mihalina, M.; Djetelic, N.

    2010-01-01

    The ultimate heat sink (UHS) is of highest importance for nuclear power plant safe and reliable operation. The most important component in line from safety-related heat sources to the ultimate heat sink water body is a component cooling heat exchanger (CC Heat Exchanger). The Component Cooling Heat Exchanger has a safety-related function to transfer the heat from the Component Cooling (CC) water system to the Service Water (SW) system. SW systems throughout the world have been the root of many plant problems because the water source, usually river, lake, sea or cooling pond, are conductive to corrosion, erosion, biofouling, debris intrusion, silt, sediment deposits, etc. At Krsko NPP, these problems usually cumulate in the summer period from July to August, with higher Sava River (service water system) temperatures. Therefore it was necessary to continuously evaluate the CC Heat Exchanger operation and confirm that the system would perform its intended function in accordance with the plant's design basis, given as a minimum heat transfer rate in the heat exchanger design specification sheet. The Essential Service Water system at Krsko NPP is an open cycle cooling system which transfers heat from safety and non-safety-related systems and components to the ultimate heat sink the Sava River. The system is continuously in operation in all modes of plant operation, including plant shutdown and refueling. However, due to the Sava River impurities and our limited abilities of the water treatment, the system is subject to fouling, sedimentation buildup, corrosion and scale formation, which could negatively impact its performance being unable to satisfy its safety related post accident heat removal function. Low temperature difference and high fluid flows make it difficult to evaluate the CC Heat Exchanger due to its specific design. The important effects noted are measurement uncertainties, nonspecific construction, high heat transfer capacity, and operational specifics (e

  17. Generating Concise Natural Language Summaries.

    McKeown, Kathleen; And Others

    1995-01-01

    Presents an approach to summarization that combines information from multiple facts into a single sentence using linguistic constructions. Describes two applications: one produces summaries of basketball games, and the other contains summaries of telephone network planning activity. Both summarize input data as opposed to full text. Discusses…

  18. Standards development status. Summary report

    1981-12-01

    The Standards Development Status Summary Report is designed for scheduling, monitoring, and controlling the process by which Regulatory Standards, Guides, Reports, Petitions, and Environmental Statements are written. It is a summary of the current schedule plans for development of the above products

  19. Statistical summary 1990-91

    1991-01-01

    The information contained in this statistical summary leaflet summarizes in bar charts or pie charts Nuclear Electric's performance in 1990-91 in the areas of finance, plant and plant operations, safety, commercial operations and manpower. It is intended that the information will provide a basis for comparison in future years. The leaflet also includes a summary of Nuclear Electric's environmental policy statement. (UK)

  20. Mineral Commodity Summaries 2009

    ,

    2009-01-01

    Each chapter of the 2009 edition of the U.S. Geological Survey (USGS) Mineral Commodity Summaries (MCS) includes information on events, trends, and issues for each mineral commodity as well as discussions and tabular presentations on domestic industry structure, Government programs, tariffs, 5-year salient statistics, and world production and resources. The MCS is the earliest comprehensive source of 2008 mineral production data for the world. More than 90 individual minerals and materials are covered by two-page synopses. For mineral commodities for which there is a Government stockpile, detailed information concerning the stockpile status is included in the two-page synopsis. Because specific information concerning committed inventory was no longer available from the Defense Logistics Agency, National Defense Stockpile Center, that information, which was included in earlier Mineral Commodity Summaries publications, has been deleted from Mineral Commodity Summaries 2009. National reserves and reserve base information for most mineral commodities found in this report, including those for the United States, are derived from a variety of sources. The ideal source of such information would be comprehensive evaluations that apply the same criteria to deposits in different geographic areas and report the results by country. In the absence of such evaluations, national reserves and reserve base estimates compiled by countries for selected mineral commodities are a primary source of national reserves and reserve base information. Lacking national assessment information by governments, sources such as academic articles, company reports, common business practice, presentations by company representatives, and trade journal articles, or a combination of these, serve as the basis for national reserves and reserve base information reported in the mineral commodity sections of this publication. A national estimate may be assembled from the following: historically reported

  1. Fuel Assembly Damping Summary

    Lee, Kanghee; Kang, Heungseok; Oh, Dongseok; Yoon, Kyungho; Kim, Hyungkyu; Kim, Jaeyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping

  2. Development of methodologies for optimization of surveillance testing and maintenance of safety related equipment at NPPs. Report of a research coordination meeting. Working material

    NONE

    1997-09-01

    This report summarizes the results of the first meeting of the Coordinated Research Programme (CRP) on Development of Methodologies for Optimization of Surveillance Testing and Maintenance of Safety Related Equipment at NPPs, held at the Agency Headquarters in Vienna, from 16 to 20 December 1996. The purpose of this Research Coordination Meeting (RCM) was that all Chief Scientific Investigators of the groups participating in the CRP presented an outline of their proposed research projects. Additionally, the participants discussed the objective, scope, work plan and information channels of the CRP in detail. Based on these presentations and discussions, the entire project plan was updated, completed and included in this report. This report represents a common agreed project work plan for the CRP. Refs, figs, tabs.

  3. Standard practice for qualification of journeyman painters for application of coatings to concrete surfaces of safety-related areas in nuclear facilities

    Anon.

    1984-01-01

    This practice provides a standard qualifying method for journeyman painters to verify their proficiency and ability to attain the required quality for application of specified coatings to concrete surfaces in safety-related areas in a nuclear facility. Variations or simplifications of the practice set forth herein may be appropriate for special coating work such as maintenance. It is not the intent of this practice to mandate a singular basis for all qualifications. Evaluation of the journeyman painter being qualified in accordance with this practice, shall be by qualified agents as specified in 4.1. Reports shall be prepared as specified in Section 5, and qualification as specified in Section 6. It is the intent of this practice to judge only the ability of the journeyman painter to apply specified coatings with the proper tools and equipment

  4. Standard practice for qualification of journeyman painters for application of coatings to steel surfaces of safety-related areas in nuclear facilities

    Anon.

    1984-01-01

    This practice provides a standard qualifying method for journeyman painters to verify their proficiency and ability to attain the required quality for application of specified coatings to steel surfaces in safety-related areas in a nuclear facility. Variations or simplifications of the practice set forth herein may be appropriate for special coating work such as maintenance or qualifications of equipment suppliers shop personnel. It is not the intent of this practice to mandate a singular basis for all qualifications. Evaluation of the journeyman painter being qualified in accordance with this practice, shall be by qualified agents as specified in 4.1. Reports shall be prepared as specified in Section 5, and certification as specified in Section 6. It is the intent of this practice to judge only the ability of the journeyman painter to apply specified coatings with the proper tools and equipment. This standard may involve hazardous materials, operations, and equipment

  5. Safety-related site characteristics - a relative comparison of the Forsmark reference areas; Saekerhetsrelaterade platsegenskaper - en relativ jaemfoerelse av Forsmark med referensomraaden

    Winberg, Anders (Conterra AB, Uppsala (Sweden))

    2010-12-15

    SKB has over the years from 2002 to 2008 conducted site investigations in Forsmark and Laxemar, with associated site modeling, design and safety analysis. In mid-2009 Forsmark was selected on the basis of analysis made as site for a future repository for spent nuclear fuel. Based on defined safety-related geoscientific location factors data from Forsmark are compared in relative terms with data from a number of locations in Sweden, previously studied by SKB. The factors compared include: the rock's composition and structures, future climate evolution, rock mechanical conditions, earthquakes, groundwater flow, groundwater composition, delay of solutes, and the ability to characterize and describe the location. Past comparisons of these properties for the selected sites show that none of these sites collectively show any significant benefit over Forsmark site for a repository. This does not preclude that there may be places on the basis of an overall assessment of geoscientific location factors could be equivalent to Forsmark

  6. Feasibility study and uncertainties in the validation of an existing safety-related control circuit with the ISO 13849-1:2006 design standard

    Jocelyn, Sabrina; Baudoin, James; Chinniah, Yuvin; Charpentier, Philippe

    2014-01-01

    In industry, machine users and people who modify or integrate equipment often have to evaluate the safety level of a safety-related control circuit that they have not necessarily designed. The modifications or integrations may involve work to make an existing machine that does not comply with normative or regulatory specifications safe. However, how can a circuit performing a safety function be validated a posteriori? Is the validation exercise feasible? What are the difficulties and limitations of such a procedure? The aim of this article is to answer these questions by presenting a validation study of a safety function of an existing machine. A plastic injection molding machine is used for this study, as well as standard ISO 13849-1:2006. Validation consists of performing an a posteriori (post-design) estimation of the performance level of the safety function. The procedure is studied for two contexts of use of the machine: in industry, and in laboratory. The calculations required by the ISO standard were done using Excel, followed by SIStema software. It is shown that, based on the context of use, the estimated performance level was different for the same safety-related circuit. The variability in the results is explained by the assumptions made by the person undertaking the validation without the involvement of the machine designer. - Highlights: • Validation of the performance level of a safety function is undertaken. • An injection molding machine and ISO 13849-1:2006 standard are used for the procedure. • The procedure is undertaken for two contexts of use of the machine. • In this study, the performance level depends on the context of use. • The assumptions made throughout the study partially explain this difference

  7. BEAUTY'99 Conference Summary

    Eerola, Paula

    2000-01-01

    Investigations of B hadrons are expected to break new ground in measuring CP-violation effects. This series of BEAUTY conferences, originating from the 1993 conference in Liblice, has contributed significantly in developing ideas of CP-violation measurements using B hadrons and formulating and comparing critically the B-physics experiments. In the '99 conference in Bled we saw the ripening of the field and the first fruit emerging - Tevatron have produced beautiful B-physics results and more are expected to come with the next run, while the B-physics experiments at DESY, SLAC and KEK are starting their operation. The longer-term projects at LHC and Tevatron have taken their shape and detailed prototyping work is going on. Meanwhile, on the phenomenological side, there has been impressive theoretical progress in understanding deeper the 'standard' measurements and proposing new signatures. In this summary, I will highlight the status of the field as presented in the conference, concentrating on signatures, experiments and R and D programmes

  8. Astronautics summary and prospects

    Kiselev, Anatoly Ivanovich; Menshikov, Valery Alexandrovich

    2003-01-01

    The monograph by A.I.Kiselev, A.A. Medvedev and Y.A.Menshikov, Astronautics: Summary and Prospects, aroused enthusiasm both among experts and the public at large. This is due to the felicitous choice of presentation that combines a simple description of complex space matters with scientificsubstantiation of the sub­ jectmatter described. The wealth of color photos makes the book still more attractive, and it was nominated for an award at the 14th International Moscow Book Fair, being singled out as the "best publication of the book fair". The book's popularity led to a second edition, substantially revised and enlarged. Since the first edition did not sufficiently cover the issues of space impact on ecology and the prospective development of space systems, the authors revised the entire volume, including in it the chapter "Space activity and ecology" and the section "Multi-function space systems". Using the federal monitoring system, now in the phase of system engi­ neering, as an example, the authors consi...

  9. Summary of the Workshop

    Myers, S; Zimmermann, F

    2012-01-01

    The summary session of the LHC Performance Workshop in Chamonix, 6-10 February 2012, synthesized one week of presentations and intense discussions on the near-, medium- and long-term strategy for the LHC and LHC upgrades. In particular, Chamonix’12 discussed the lessons from 2011, the strategy, beam energy and beam parameters for 2012, the planning for the Long Shutdown no. 1 (LS1), the measures and schemes for improving or maintaining the machine availability at higher beam energy, the injector performance and injector upgrade schedule, the HL-LHC project as well as possible additional or future LHC upgrades like LHeC and HELHC. Key workshop themes included the risk associated with 4 TeV beam energy in 2012, the beam energy after LS1, the turnaround time, the physics goal and optimized running schedule for 2012, the achievements and plans for Pb-Pb and p-Pb collisions, beam-beam effects, electron-cloud phenomena and UFOs. We report the proposals for decisions which have emerged at the Chamonix’12 workshop. (author)

  10. Mineral commodity summaries 2013

    ,

    2013-01-01

    Each chapter of the 2013 edition of the U.S. Geological Survey (USGS) Mineral Commodity Summaries (MCS) includes information on events, trends, and issues for each mineral commodity as well as discussions and tabular presentations on domestic industry structure, Government programs, tariffs, 5-year salient statistics, and world production and resources. The MCS is the earliest comprehensive source of 2012 mineral production data for the world. More than 90 individual minerals and materials are covered by two-page synopses. For mineral commodities for which there is a Government stockpile, detailed information concerning the stockpile status is included in the two-page synopsis. Abbreviations and units of measure, and definitions of selected terms used in the report, are in Appendix A and Appendix B, respectively. “Appendix C—Reserves and Resources” includes “Part A—Resource/Reserve Classification for Minerals” and “Part B—Sources of Reserves Data.” A directory of USGS minerals information country specialists and their responsibilities is Appendix D. The USGS continually strives to improve the value of its publications to users. Constructive comments and suggestions by readers of the MCS 2013 are welcomed.

  11. Mineral commodity summaries 2014

    ,

    2014-01-01

    Each chapter of the 2014 edition of the U.S. Geological Survey (USGS) Mineral Commodity Summaries (MCS) includes information on events, trends, and issues for each mineral commodity as well as discussions and tabular presentations on domestic industry structure, Government programs, tariffs, 5-year salient statistics, and world production and resources. The MCS is the earliest comprehensive source of 2013 mineral production data for the world. More than 90 individual minerals and materials are covered by two-page synopses. For mineral commodities for which there is a Government stockpile, detailed information concerning the stockpile status is included in the two-page synopsis. Abbreviations and units of measure, and definitions of selected terms used in the report, are in Appendix A and Appendix B, respectively. “Appendix C—Reserves and Resources” includes “Part A—Resource/Reserve Classification for Minerals” and “Part B—Sources of Reserves Data.” A directory of USGS minerals information country specialists and their responsibilities is Appendix D. The USGS continually strives to improve the value of its publications to users. Constructive comments and suggestions by readers of the MCS 2014 are welcomed.

  12. Summary of the Workshop

    Myers, S; Zimmermann, F [European Organization for Nuclear Research, Geneva (Switzerland)

    2012-07-01

    The summary session of the LHC Performance Workshop in Chamonix, 6-10 February 2012, synthesized one week of presentations and intense discussions on the near-, medium- and long-term strategy for the LHC and LHC upgrades. In particular, Chamonix’12 discussed the lessons from 2011, the strategy, beam energy and beam parameters for 2012, the planning for the Long Shutdown no. 1 (LS1), the measures and schemes for improving or maintaining the machine availability at higher beam energy, the injector performance and injector upgrade schedule, the HL-LHC project as well as possible additional or future LHC upgrades like LHeC and HELHC. Key workshop themes included the risk associated with 4 TeV beam energy in 2012, the beam energy after LS1, the turnaround time, the physics goal and optimized running schedule for 2012, the achievements and plans for Pb-Pb and p-Pb collisions, beam-beam effects, electron-cloud phenomena and UFOs. We report the proposals for decisions which have emerged at the Chamonix’12 workshop. (author)

  13. INTRODUCTION Summary of Papers Summary of Papers

    Gauthier, Serge; Abarzhi, Snezhana I.; Sreenivasan, Katepalli R.

    2010-12-01

    passive scalar, a velocity component, turbulent kinetic energy and dissipation rate. The analyses of the DNS data for homogeneous shear flows show that statistically the gradient vectors with large magnitudes align with each other, while gradients with small magnitudes tend to be randomly organized. Zybin et al study turbulence structure through a model of vortex filament. In this way, they show that contraction and stretching out of a filament provide an energy flux from larger to smaller scales. The authors obtain the scaling exponents for both Lagrangian and transverse Eulerian structure functions and report good agreement with the existing data. Wall-bounded flows. Six papers are focused on the theme of wall-bounded flows. Cassel and Obabko perform numerical simulations of the two-dimensional flow induced by a thick-core vortex. This problem is important for studies of unsteady separation in the vortex-induced flows. Their accurate investigations convicingly justify that the Rayleigh instability does exist at large Reynolds numbers. Cvitanović and Gibson study the effects of geometry on transitional turbulent flow and focus on wall-bounded shear flows at moderate Reynolds numbers. The authors determine a set of unstable periodic orbits from close recurrences of the turbulent flow, identify a few equilibria that resemble frequently observed but unstable coherent structures, and construct a low-dimensional state-space projection from the extremely high-dimensional data sets. The approach developed by the authors can be a useful tool for understanding massive data sets. Seidel et al focus on developing feedback flow control strategies, i.e., they attempt to achieve a desired flow state for the turbulent shear layer behind a backward facing step. The authors show that the Proper Orthogonal Decomposition (POD) of the density field is a better marker than that for the velocity field, as in the former case the contribution of small scale structures is effectively eliminated

  14. Summary cortisol reactivity indicators: Interrelations and meaning

    Jennifer E. Khoury

    2015-01-01

    Full Text Available Research on the hypothalamic pituitary adrenal (HPA axis has involved a proliferation of cortisol indices. We surveyed recently published HPA-related articles and identified 15 such indices. We sought to clarify their biometric properties, specifically, how they interrelate and what they mean, because such information is rarely offered in the articles themselves. In the present article, the primary samples consist of community mothers and their infants (N = 297, who participated in two challenges, the Toy Frustration Paradigm and the Strange Situation Procedure. We sought to cross-validate findings from each of these samples against the other, and also against a clinically depressed sample (N = 48 and a sample of healthy older adults (N = 51 who participated in the Trier Social Stress Test. Cortisol was collected from all participants once before and twice after the challenges. These heterogenous samples were chosen to obtain the greatest possible range in cortisol levels and stress response regulation. Using these data, we computed the 15 summary cortisol indices identified in our literature survey. We assessed inter-relations amongst indices and determined their underlying dimensions via principal component analysis (PCA. The PCAs consistently extracted two components, accounting for 79%–93% of the variance. These components represent “total cortisol production” and “change in cortisol levels.” The components were highly congruent across challenge, time, and sample. High variable loadings and explained factor variance suggest that all indices represent their underlying dimensions very well. Thus the abundance of summary cortisol indices currently represented in the literature appears superfluous.

  15. How components are qualified

    Burstein, N.M.

    1980-01-01

    This work presents a basic description and rationale behind the requirements, reasons and implementation of the concept of qualification for Safety Related Class 1E electric equipment for use in Nuclear Power Generating Stations. The discussion centers around the requirements as set forth in the Code of Federal Regulations, IEEE Standard 323, and the guidance provided by Regulatory Guide 1.89. The main elements of the study describe what is meant by qualification, what is required for qualification and the implementation process. 11 refs

  16. Photonics: Technology project summary

    Depaula, Ramon P.

    1991-01-01

    Photonics involves the use of light (photons) in conjunction with electronics for applications in communications, computing, control, and sensing. Components used in photonic systems include lasers, optical detectors, optical wave guide devices, fiber optics, and traditional electronic devices. The goal of this program is to develop hybrid optoelectronic devices and systems for sensing, information processing, communications, and control. It is hoped that these new devices will yield at least an order of magnitude improvement in performance over existing technology. The objective of the program is to conduct research and development in the following areas: (1) materials and devices; (2) networking and computing; (3) optical processing/advanced pattern recognition; and (4) sensing.

  17. Summary of the Day (CDMP)

    National Oceanic and Atmospheric Administration, Department of Commerce — This Summary of the Day data file contains daily selected elements of observations recorded by certified observers. The stations were located in the U.S. and were...

  18. Long term performance session summary

    Hanauer, S.

    1996-05-01

    This paper presents brief summaries of reports given on plutonium disposal. Topics include: performance of waste forms; glass leaching; ceramic leaching; safeguards and security issues; safeguards of vitrification; and proliferation risks of geologic disposal.

  19. Summary of Meson'98 Workshop

    Henley, E.M.

    1998-01-01

    One never quite knows what to say in a summary. If you were at the sessions, you heard the same talks I did. Perhaps the purpose is to summarize the parallel sessions, but like you, I can only attend one of these sessions. In addition, the time is short, so that this cannot be a real summary. What I will present are impressions of the past two days, and these will certainly be colored by my own views. Thus at the outset, let me apologize for any and all omissions and distortions. I will cover primarily the plenary session talks, but will organize this summary along the following lines: 1. vector (V) mesons; 2. pseudoscalar mesons, and 3. other subjects, notably with electrons. This afternoon's talks are so close in time to this summary that I shall omit them. (author)

  20. Operating reactors licensing actions summary

    1981-08-01

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  1. Summary 2010 Greenhouse Gas Data

    U.S. Environmental Protection Agency — This file contains a summary of the publicly available data from the GHG Reporting Program for 2010. This data includes non-confidential data reported by facilities...

  2. Accelerator technology working group summary

    Jameson, R.A.

    1985-01-01

    A summary is presented of workshop deliberations on basic scaling, the economic viability of laser drive power for HEP accelerators, the availability of electron beam injectors for near-term experiments, and a few very general remarks on technology issues

  3. Summary and conclusions [Chapter 11

    Daniel G. Neary; John N. Rinne; Alvin L.. Medina

    2012-01-01

    Summaries and conclusions of each chapter are compiled here to provide a “Quick Reference” guide of major results and recommendations for the UVR. More detail can be obtained from individual chapters.

  4. Influence of Non-safety Important Component on Maintenance Rule

    Ju, Tae Young; Kim, Wang Bae

    2016-01-01

    The Maintenance Rule (MR) programs in KHNP have been implemented since Jan 2009. KHNP is currently developing MR program for new built plant which has been constructed from December 2011. It is required to utilize plant-specific probabilistic safety analysis (PSA) result as risk significant criteria to determine which components are significantly important to safety. The criteria consist of three PSA risk values which are risk reduction worth (RRW), risk achievement worth (RAW) and core damage frequency (CDF) contribution. Most safety related components are classified as high risk significant, and non-safety related components as low safety significant in MR program. This paper presents the influence of the non-safety related component which has high PSA risk value on MR program of new built plant. It is considered that safety related system has at least one or more safety functions and some non-safety functions, but non-safety system doesn't have any safety function. The safety functions are defined as three functions which are required to maintain 1) integrity of reactor coolant pressure boundary, 2) capability to shut-down the reactor and maintain it in a safe shutdown, and 3) capability to prevent or mitigate the accident that could result in potential offsite exposure. The Maintenance Rule program is developed based on PSA result. Safety functions have high risk value in PSA program and considered HSS function in MR program. On the contrary, non-safety functions are generally has low risk value in PSA program and they are determined as LSS function in MR program. The AAC DG and its supporting systems are designed as non-safety systems which mean they don't have any safety function. But, AAC DG is treated as an important measure to mitigate accident in PSA program. It is determined as HSS function in MR program because it has high risk value in PSA program. AAC DG supporting systems does not have high risk value in operating plant's PSA program

  5. Safeguards Summary Event List (SSEL)

    1984-03-01

    The Safeguards Summary Event List (SSEL) provides brief summaries of several hundred safeguards-related events involving nuclear material or facilities regulated by the U.S. Nuclear Regulatory Commission (NRC). Events are described under the categories of bomb-related, intrusion, missing and/or allegedly stolen, transportation, tampering/vandalism, arson, firearms, radiological sabotage and miscellaneous. The information contained in the event descriptions is derived primarily from official NRC reporting channels

  6. Safeguards Summary Event List (SSEL)

    1983-02-01

    The Safeguards Summary Event List (SSEL) provides brief summaries of several hundred safeguards-related events involving nuclear material or facilities regulated by the US Nuclear Regulatory Commission (NRC). Events are described under the categories of bomb-related, intrusion, missing and/or allegedly stolen, transportation, vandalism, arson, firearms, radiological sabotage and miscellaneous. The information contained in the event descriptions is derived primarily from official NRC reporting channels

  7. Safeguards Summary Event List (SSEL)

    1982-07-01

    The Safeguards Summary Event List (SSEL) provides brief summaries of several hundred safeguards-related events involving nuclear material or facilities regulated by the US Nuclear Regulatory Commission (NRC). Events are described under the categories of bomb-related, intrusion, missing and/or allegedly stolen, transportation, vandalism, arson, firearms, sabotage and miscellaneous. The information contained in the event descriptions is derived primarily from official NRC reporting channels

  8. Improvement of risk informed surveillance test interval for the safety related instrument and control system of Ulchin units 3 and 4

    Jang, Seung Cheol; Lee, Yun Hwan; Lee, Seung Joon; Han, Sang Hoon

    2012-05-01

    The purpose of this research is the development of various methodologies necessary for the licensing of the risk informed surveillance test interval(STI) improvement for the safety related I and C systems in UCN 3 and 4, for instance, reactor protection system (RPS), engineered safety features actuation system (ESFAS), ESF auxiliary relay cabinet (ARC), and core protection calculator (CPC). The technical adequacy of the methodology was sufficiently verified through the application to the following STI changes. o CPC channel functional test (change from 1 month to 3 months including safety channel and log power test) o RPS channel functional test (change from 1 month to 3 months) o RPS logic and trip channel test (change from 1 month to 3 months. 1 month for RPS manual actuation test) o ESFAS channel functional test (change from 1 month to 3 months) o ESFAS logic and trip channel test (change from 1 month to 3 months) o ESF auxiliary relay test (change from 1 month to 3 months with staggered test. Manual actuation at the ESF ARC is added as a backup of ESF actuation signals during emergency operation

  9. Improvement of risk informed surveillance test interval for the safety related instrumentation and control system of Yonggwang units 3 and 4

    Jang, Seung Cheol; Lee, Yun Hwan; Lee, Seung Joon; Han, Sang Hoon

    2012-05-01

    The purpose of this research is the development of various methodologies necessary for the licensing of the risk informed surveillance test interval(STI) improvement for the safety related I and C systems in YGN 3 and 4, for instance, reactor protection system (RPS), engineered safety features actuation system (ESFAS), ESF auxiliary relay cabinet (ARC), and core protection calculator (CPC). The technical adequacy of the methodology was sufficiently verified through the application to the following STI changes. o CPC channel functional test (change from 1 month to 3 months including safety channel and log power test) o RPS channel functional test (change from 1 month to 3 months) o RPS logic and trip channel test (change from 1 month to 3 months. 1 month for RPS manual actuation test) o ESFAS channel functional test (change from 1 month to 3 months) o ESFAS logic and trip channel test (change from 1 month to 3 months) o ESF auxiliary relay test (change from 1 month to 3 months with staggered test. Manual actuation at the ESF ARC is added as a backup of ESF actuation signals during emergency operation

  10. Conceptual and safety-related questions in the final storage of radioactive waste - a comparison of various types of host rock

    Kleemann, U.

    2005-01-01

    The German Federal Office for Radiation Protection (BfS) in early November published the synthesis report (BfS 2005) about the conceptual and safety-related specific questions associated with the final storage of radioactive waste. In addition to a condensed version of twelve individual projects, the report contains a description of the results of the peer review and the workshops carried out, in particular an evaluation comparing different types of host rock in Germany. The whole project constitutes a comprehensive documentation of the current state of the art. Findings are expressed at a general level referring neither to the suitability of any specific repository site nor to that of salts as a repository formation, but covering all potential repository formations in deep geologic strata in Germany. The limits to and possibilities of, generic comparisons of various types of host rock are shown. It si seen that, in principle, none of the host rock varieties in Germany would be preferable to others. Numerous problems can be solved only for specific sites, thus requiring site comparisons. While some questions indicate a need for regulatory treatment, the need for basic research is considered to be low. The contribution presents the main findings made in each of the specific projects and the evaluations by the Office. (orig.)

  11. Requirements to be taken into account in the design, qualification startup and operation of electrical equipment for safety-related electrical systems

    1985-07-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The purpose of this RFS is to provide the rules to be respected in order that safety-related electrical systems can perform its function under plausible operating conditions

  12. Steel-plate composite (SC) walls for safety related nuclear facilities: Design for in-plane forces and out-of-plane moments

    Varma, Amit H.; Malushte, Sanjeev R.; Sener, Kadir C.; Lai, Zhichao

    2014-01-01

    Steel-concrete (SC) composite walls being considered and used as an alternative to conventional reinforced concrete (RC) walls in safety-related nuclear facilities due to their construction economy and structural efficiency. However, there is a lack of standardized codes for SC structures, and design guidelines and approaches are still being developed. This paper presents the development and verification of: (a) mechanics based model, and (b) detailed nonlinear finite element model for predicting the behavior and failure of SC wall panels subjected to combinations of in-plane forces. The models are verified using existing test results, and the verified models are used to explore the behavior of SC walls subjected to combinations of in-plane forces and moments. The results from these investigations are used to develop an interaction surface in principle force (S p1 –S p2 ) space that can be used to design or check the adequacy of SC wall panels. The interaction surface is easy to develop since it consists of straight line segments connecting anchor points defined by the SC wall section strengths in axial tension, in-plane shear, and compression. Both models and the interaction surface (for design) developed in this paper are recommended for future work. However, in order to use these approaches, the SC wall section should be detailed with adequate shear connector and tie bar strength and spacing to prevent non-ductile failure modes

  13. 7 CFR 3402.12 - Project summary.

    2010-01-01

    ... 7 Agriculture 15 2010-01-01 2010-01-01 false Project summary. 3402.12 Section 3402.12 Agriculture... FELLOWSHIP GRANTS PROGRAM Preparation of an Application § 3402.12 Project summary. Using the Project Summary.... The summary should not include any reference to the specific number of fellowships requested. The...

  14. 49 CFR 194.113 - Information summary.

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Information summary. 194.113 Section 194.113... Response Plans § 194.113 Information summary. (a) The information summary for the core plan, required by... state(s). (b) The information summary for the response zone appendix, required in § 194.107, must...

  15. 2011 Annual Criticality Safety Program Performance Summary

    Andrea Hoffman

    2011-12-01

    The 2011 review of the INL Criticality Safety Program has determined that the program is robust and effective. The review was prepared for, and fulfills Contract Data Requirements List (CDRL) item H.20, 'Annual Criticality Safety Program performance summary that includes the status of assessments, issues, corrective actions, infractions, requirements management, training, and programmatic support.' This performance summary addresses the status of these important elements of the INL Criticality Safety Program. Assessments - Assessments in 2011 were planned and scheduled. The scheduled assessments included a Criticality Safety Program Effectiveness Review, Criticality Control Area Inspections, a Protection of Controlled Unclassified Information Inspection, an Assessment of Criticality Safety SQA, and this management assessment of the Criticality Safety Program. All of the assessments were completed with the exception of the 'Effectiveness Review' for SSPSF, which was delayed due to emerging work. Although minor issues were identified in the assessments, no issues or combination of issues indicated that the INL Criticality Safety Program was ineffective. The identification of issues demonstrates the importance of an assessment program to the overall health and effectiveness of the INL Criticality Safety Program. Issues and Corrective Actions - There are relatively few criticality safety related issues in the Laboratory ICAMS system. Most were identified by Criticality Safety Program assessments. No issues indicate ineffectiveness in the INL Criticality Safety Program. All of the issues are being worked and there are no imminent criticality concerns. Infractions - There was one criticality safety related violation in 2011. On January 18, 2011, it was discovered that a fuel plate bundle in the Nuclear Materials Inspection and Storage (NMIS) facility exceeded the fissionable mass limit, resulting in a technical safety requirement (TSR) violation. The

  16. Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-category 1 systems for the San Onofre Nuclear Power Plant, Unit 1

    Latorre, V.R.; Victor, R.A.

    1980-11-01

    This report documents the technical evaluation of Southern California Edison Company's San Onofre Nuclear Power Plant, Unit 1, to determine whether the failure of any non-Category 1 (seismic) equipment could result in a condition, such as flooding, that might potentially adversely affect the performance of safety-related equipment required for the safe shutdown of the facility or to mitigate the consequences of an accident. Criteria developed by the US Nuclear Regulatory Commission were used to evaluate the acceptability of the existing protection as well as measures taken by Southern California Edison Company to minimize the danger of flooding and to protect safety-related equipment

  17. Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-category 1 systems for the Yankee Rowe Nuclear Power Station

    Epps, R.C.

    1980-11-01

    This report documents the technical evaluation of the Maine Yankee Atomic Power Station. The purpose of this evaluation was to determine whether the failure of any non-Class I (seismic) equipment could result in a condition, such as flooding, that might adversely affect the performance of the safety-related equipment required for the safe shutdown of the facility, or to mitigate the consequences of an accident. Criteria developed by the US Nuclear Regulatory Commission were used to evaluate the acceptability of the existing protection system as well as measures taken by Maine Yankee Atomic Power Company (MYAPC) to minimize the danger of flooding and to protect safety-related equipment

  18. Grain alcohol study: summary

    The study has concentrated upon a detailed examination of all considerations involved in the production, use, and marketing of ethyl alcohol (ethanol) as produced from the fermentation of agricultural grains. Each parameter was examined in the light of current energy markets and trends; new sources and technological, and processes for fermentation, the capability of the agricultural industry to support fermentation demand; the optimizaton of value of agricultural crops; and the efficiencies of combining related industries. Ahydrous (200 proof) ethanol makes an excellent blending component for all present automotive fuels and an excellent octane additive for unleaded fuels in proportions up to 35% without requiring modifications to current engines. There is no difference between ethanol produced by fermentation and ethanol produced synthetically from petroleum. The decision to produce ethanol one way or the other is purely economic. The agricultural industry can support a major expansion in the fermentation industry. The residue (distillers grains) from the fermentation of corn for ethanol is an excellent and economical feed for livestock and poultry. A reliable supply of distillers grain can assist in making the large beef feedlot operations more economically viable. The source materials, fuels, products and by-products of an ethanol plant, beef feedlot, gas biodigester plant, municipal waste recovery plant and a steam generated electrical plant are interrelated and mutually beneficial for energy efficiencies and economic gains when co-located. The study concludes that the establishment of such agricultural- environment industrial energy complexes, would provide a broad range of significant benefits to Indiana.

  19. Grain alcohol study: summary

    The study has concentrated upon a detailed examination of all considerations involved in the production, use, and marketing of ethyl alcohol (Ethanol) as produced from the fermentation of agricultural grains. Each parameter was examined in the light of current energy markets and trends; new sources and technological, and processes for fermentation, the capability of the agricultural industry to support fermentaton demand; the optimization of value of agricultureal crops; and the efficiencies of combining related industries. Anhydrous (200 proof) ethanol makes an excellent blending component for all present automotive fuels and an excellent octane additive for unleaded fuels in proportions up to 35% without requiring modifications to current engines. There is no difference between ethanol produced by fermentation and ethanol produced synthetically from petroleum. The decision to produce ethanol one way or the other is purely economic. The agricultural industry can support a major expansion in the fermentation industry. The residue (distillers grains) from the fermentation of corn for ethanol is an excellent and economical feed for livestock and poultry. A reliable supply of distillers grains can assist in making the large beef feedlot operations more economically viable. The source materials, fuels, products and by-products of an ethanol plant, beef feedlot, gas biodigester plant, municipal waste recovery plant and a steam generated electrical plant are interrelated and mutually beneficial for energy efficiencies and economic gains when co-located. The study concludes that the establishment of such agricultural-environment industrial energy complexes, would provide a broad range of significant benefits to Indiana.

  20. Overview and summary remarks

    Haynes, R.H.

    1990-01-01

    The non-scientific community believes that there is such a thing as safety and that this should be an attainable goal in any technology. People are poor intuitive judges of risks. If the first inducer of mutation to be discovered had been mustard gas rather than ionizing radiation, we probably would not be as concerned about radiation. Radiation is a relatively poor mutagen; it is a good recombinogen but poor compared to ultraviolet and may chemicals for the production of point mutagens. The issue of threshold doses has not been resolved; when one allows for the existence of repair, one opens the possibility that repair-proficient cells could have a dose-response curve with a zero slope at very low doses. Another issue concerns mutagen burden, particularly when comparing radiation with chemicals. Exposure to ionizing radiation may be a very minor component of the total mutagen burden to which we are all exposed. There can exist both synergistic and antagonistic interactions among mutagens. We should not forget the role of metabolism in genetic responses. The responses of cells to mutagens of all kinds are much more complex biochemically than one would imagine on the basis of the simple notion of DNA damage and its error-free or error-prone repair. (L.L.)

  1. Executive Summary - Overview

    2005-01-01

    Professor Andrzej Budzanowski was Director of IFJ in the years 1990-2004. On September 1st 2004, on IFJ joining the Polish Academy of Sciences, Professor Marek Jezabek has been nominated by the President of PAN as the Director of IFJ, for a 4-year term. Our Institute, with a personnel of 450 (182 research staff) and over 50 Ph.D. students, is presently one of the largest institutes of the Polish Academy of Sciences and one of the largest research institutes in Poland. The scientific staff consists on 120 post-doctoral researchers, 26 Associated Professors and 36 State-Nominated Professors. The total budget of the Institute for the year 2004 was about 56 million Euro. The IFJ is financed mainly from the state budget of the Ministry of Scientific Research and Information Technology. In 2004 this financing was about 4 million Euro, constituting 72% of the Institute's total budget. The remaining part of our 2004 budget came from individual research projects, also sponsored by the Ministry of Scientific Research (688 kEuro), from international projects (368 kEuro) and from the Institute's entrepreneurship activities (498 kEuro). Between 2003 and 2004 we doubled our income from international projects. For further budget information. The Scientific Council of the Institute, which consists of 40 elected members of the Institute's staff and 4 external members (elected representatives from other Polish institutes and universities), is authorized to confer Ph.D. degrees in Physics and related disciplines, and to initiate and conduct habilitation and professorship procedures. In 2003-2004 17 Ph.D. theses and 11 habilitations have been completed. Following their review procedures, 3 Associate Professors at the IFJ became state-nominated Professors, receiving their nominations from Poland's President, Mr. A. Kwasniewski. The Institute is structured into 17 scientific departments which cover the range of our scientific interests. A summary of our main scientific achievements in

  2. GERB viscous dampers in applications for pipelines and other components in Czechoslovak nuclear power plants

    Masopust, R.; Podrouzek, J.

    1992-01-01

    VISCODAMPERS from GERB, Germany, are now widely used as reliable shock restraints against earthquake and other shock effects for the most important safety-related pipelines and components in several Czechoslovak nuclear power plants. Having many technical advantages they are, at the same time, relatively inexpensive in comparison with conventional snubbers. Their properties are briefly described and several practical applications are explained. (author) 3 tabs., 9 figs., 8 refs

  3. GERB viscous dampers in application for pipelines and other components in nuclear power plants

    Masopust, R.; Podrouzek, J.; Zach, J.

    1993-01-01

    VISCODAMPERS from GERB, Germany, are now widely used as reliable shock restraints against earthquake and other shock effects for the most important safety-related pipelines and components in several Czech and Slovak nuclear power plants. Having many technical advantages they are, at the same time, relatively inexpensive in comparison to conventionally used snubbers. Their properties are briefly described and several practical applications are explained in this paper. (author)

  4. HEAO Block 2 study executive summary

    1976-03-01

    An executive summary is presented of a preliminary study done on several potential High Energy Astronomy Observatory (HEAO) missions which are follow-on missions to the currently defined HEAO program. The purpose was to examine several typical missions and determine the relative complexities associated with them. The four payloads investigated were a 1.2 m Diameter x-ray Telescope observatory, a Large Area Moderate Angular Resolution (LAMAR) observatory, a cosmic ray observatory, and (4) a gamma ray observatory. Each of the four observatories was considered a national facility. Low cost approaches were stressed throughout, with considerable use of HEAO Block I experience and designs effected to provide a high degree of confidence that such approaches were achievable. The use of the Multi-Mission Spacecraft (MMS) and the HEAO Block I spacecraft was considered as a result of this low cost emphasis. Also, NASA standard components were considered, where applicable

  5. Safeguards summary event list (SSEL)

    1989-07-01

    The Safeguards Summary Event List (SSEL) provides brief summaries of several hundred safeguards-related events involving nuclear material or facilities regulated by the US Nuclear Regulatory Commission (NRC). Because of public interest, also included are events reported involving byproduct material which is exempt from safeguards requirements. Events are described under the categories of bomb-related, intrusion, missing and/or allegedly stolen, transportation, tampering/vandalism, arson, firearms, radiological sabotage, nonradiological sabotage, alcohol and drugs, and miscellaneous. The information contained in the event descriptions is derived primarily from official NRC reporting channels

  6. Meson 2000 Conference Summary lite

    Barnes, T.

    2000-01-01

    This short contribution is a late MESON2000 conference summary. As appropriate for the 600th anniversary of the Jagiellonian University, it begins with a brief summary of the last 600 years of European history and its place in hadron physics. Next a ''physicist chirality'' order parameter PC is introduced. When applied to MESON2000 plenary speakers this order parameter illustrates the separation of hadron physicists into disjoint communities. The individual plenary talks in MESON2000 are next sorted according to the subconference associated with each of the 36 plenary speakers. Finally, I conclude with a previously unreported Feynman story regarding the use of models in hadron physics. (author)

  7. Operating reactors licensing actions summary

    1983-01-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  8. Safeguards Summary Event List (SSEL)

    Fadden, M.; Yardumian, J.

    1993-07-01

    The Safeguards Summary Event List provides brief summaries of hundreds of safeguards-related events involving nuclear material or facilities regulated by the US Nuclear Regulatory Commission. Events are described under the categories: Bomb-related, Intrusion, Missing/Allegedly Stolen, Transportation-related, Tampering/Vandalism, Arson, Firearms-related, Radiological Sabotage, Non-radiological Sabotage, and Miscellaneous. Because of the public interest, the Miscellaneous category also includes events reported involving source material, byproduct material, and natural uranium, which are exempt from safeguards requirements. Information in the event descriptions was obtained from official NRC sources

  9. Experimental Plasma Research project summaries

    1980-09-01

    This report contains descriptions of the activities supported by the Experimental Plasma Research Branch of APP. The individual project summaries were prepared by the principal investigators and include objectives and milestones for each project. The projects are arranged in six research categories: Plasma Properties; Plasma Heating; Plasma Diagnostics; Atomic, Molecular and Nuclear Physics; Advanced Superconducting Materials; and the Fusion Plasma Research Facility (FPRF). Each category is introduced with a statement of objectives and recent progress and followed by descriptions of individual projects. An overall budget summary is provided at the beginning of the report

  10. Operating reactors licensing actions summary

    1982-05-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  11. Operating reactors licensing actions summary

    1983-03-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  12. Operating reactors licensing actions summary

    1982-07-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  13. Operating reactors licensing actions summary

    1982-11-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  14. Experimental plasma research project summaries

    1978-08-01

    This report contans descriptions of the activities supported by the Experimental Plasma Research Branch of APP. The individual project summaries were prepared by the principal investigators and include objectives and milestones for each project. The projects are arranged in six research categories: Plasma Properties; Plasma Heating; Plasma Measurements and Instrumentation; Atomic, Molecular and Nuclear Physics; Advanced Superconducting Materials; and the Fusion Plasma Research Facility (FPRF). Each category is introduced with a statement of objectives and recent progress and followed by descriptions of individual projects. An overall budget summary is provided at the beginning of the report

  15. Experimental Plasma Research project summaries

    None

    1980-09-01

    This report contains descriptions of the activities supported by the Experimental Plasma Research Branch of APP. The individual project summaries were prepared by the principal investigators and include objectives and milestones for each project. The projects are arranged in six research categories: Plasma Properties; Plasma Heating; Plasma Diagnostics; Atomic, Molecular and Nuclear Physics; Advanced Superconducting Materials; and the Fusion Plasma Research Facility (FPRF). Each category is introduced with a statement of objectives and recent progress and followed by descriptions of individual projects. An overall budget summary is provided at the beginning of the report.

  16. Operating reactors licensing actions summary

    1982-10-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  17. Operating reactors licensing actions summary

    1982-08-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  18. Operating reactors licensing actions summary

    1982-09-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  19. Qualification of engine-mounted components due to operational vibration

    Lee, B.J.; Bayat, A.

    1994-01-01

    The Emergency Diesel Generator (EDG) in a Nuclear Power Plant is considered to be an essential component of the plant for its safe operation. Failures of auxiliary components directly mounted on the EDG creates costly repairs, and compromises the engine's availability and reliability. Although IEEE-323 and Section III of the ASME code require addressing of safety-related components due to mechanically induced vibration, very few guidelines exist in the nuclear industry to show how this may be accounted for. Most engine vendors rely on the empirical experience data as the basis of their evaluation for vibration. Upgrade of engine controls, addition of monitoring components and other engine modifications require design and installation of new mechanical and electrical components to be mounted directly on the engine. This necessitates the evaluation of such components for engine-induced vibration which is considered to be one of the most severe design parameters. This paper presents a methodology to evaluate three categories of components; structural, mechanical, and electrical under engine vibration. The discussion for the characteristics and manipulation of engine vibration profile to be used for each component evaluation is also given. In addition, the suitability of analytical verses testing approaches is discussed for each category. An example application of the methodology is presented for a typical EDG which is currently undergoing major controls upgrade and monitoring modification

  20. Improving safety-related knowledge, attitude and practices of nurses handling cytotoxic anticancer drug: pharmacists' experience in a general hospital, Malaysia.

    Keat, Chan Huan; Sooaid, Nor Suhada; Yun, Cheng Yi; Sriraman, Malathi

    2013-01-01

    An increasing trend of cytotoxic drug use, mainly in cancer treatment, has increased the occupational exposure among the nurses. This study aimed to assess the change of nurses' safety-related knowledge as well as attitude levels and subsequently to assess the change of cytotoxic drug handling practices in wards after a series of pharmacist-based interventions. This prospective interventional study with a before and after design requested a single group of 96 nurses in 15 wards actively providing chemotherapy to answer a self-administered questionnaire. A performance checklist was then used to determine the compliance of all these wards with the recommended safety measures. The first and second assessments took 2 months respectively with a 9-month intervention period. Pharmacist-based interventions included a series of technical, educational and administrative support measures consisting of the initiation of closed-system cytotoxic drug reconstitution (CDR) services, courses, training workshops and guideline updates. The mean age of nurses was 32.2∓6.19 years. Most of them were female (93.8%) and married (72.9%). The mean knowledge score of nurses was significantly increased from 45.5∓10.52 to 73.4∓8.88 out of 100 (p<0.001) at the end of the second assessment. Overall, the mean practice score among the wards was improved from 7.6∓5.51 to 15.3∓2.55 out of 20 (p<0.001). The pharmacist-based interventions improved the knowledge, attitude and safe practices of nurses in cytotoxic drug handling. Further assessment may help to confirm the sustainability of the improved practices.

  1. Regulatory standpoints on the design-basis capability of safety-related motor-operated valves(MOVs) and power-operated gate valves(POGVs)

    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

  2. Mitigating component performance variation

    Gara, Alan G.; Sylvester, Steve S.; Eastep, Jonathan M.; Nagappan, Ramkumar; Cantalupo, Christopher M.

    2018-01-09

    Apparatus and methods may provide for characterizing a plurality of similar components of a distributed computing system based on a maximum safe operation level associated with each component and storing characterization data in a database and allocating non-uniform power to each similar component based at least in part on the characterization data in the database to substantially equalize performance of the components.

  3. Biomass energy systems program summary

    None

    1980-07-01

    Research programs in biomass which were funded by the US DOE during fiscal year 1978 are listed in this program summary. The conversion technologies and their applications have been grouped into program elements according to the time frame in which they are expected to enter the commercial market. (DMC)

  4. Operating reactors licensing actions summary

    1982-04-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis

  5. Summary Staging Manual 2000 - SEER

    Access this manual of codes and coding instructions for the summary stage field for cases diagnosed 2001-2017. 2000 version applies to every anatomic site. It uses all information in the medical record. Also called General Staging, California Staging, and SEER Staging.

  6. Summary and Outlook

    Wuest, Martin; Robinson, David W.; Decoste, Dennis

    are sensitive to degradation as a function of contamination and the amount of extracted charge. Windowless electron multipliers are therefore not very stable reference detectors. This makes it difficult to obtain a reliable absolute calibration traceable to a national measurement institute. Calibration is still a time consuming process. It involves testing the instrument at component, subsystem and integrated level. It is important that the instrument is not only operated using a special calibration configuration to save time, but also in its full flight configuration exercising the full path of the data through data compression and telemetry. Very seldom there is enough time available to calibrate all the desired points in parameter space. Usually only a subset can be calibrated for schedule and economic reasons. The number of calibration points is often further reduced since the available calibration time is cut due to development schedule slip and a fixed launch date. This increases the uncertainties as more parameters have to be interpolated or extrapolated. Calibration data should be evaluated preferably in near-real time to prevent losing valuable calibration time if something in the instrument or facility is not working properly. Computer simulation models should be used to obtain a thorough understanding of the actual flight instrument. In flight the instrument performance degrades due to contamination (outgassing), environmental effects (atomic oxygen, radiation) or aging. One of the most sensitive parts in today's instrument are their detectors. Microchannel plate detectors degrade as function of the extracted charge. Solid-state detectors experience radiation damage which increases their noise and the lower energy detection threshold. The goal of the in-flight calibration is to determine this instrument degradation. Calibration is then performed by comparing measurements taken with different bias voltage or discriminator threshold settings. If possible, the

  7. A summary of demand response in electricity markets

    Albadi, M.H.; El-Saadany, E.F.

    2008-01-01

    This paper presents a summary of demand response (DR) in deregulated electricity markets. The definition and the classification of DR as well as potential benefits and associated cost components are presented. In addition, the most common indices used for DR measurement and evaluation are highlighted, and some utilities' experiences with different demand response programs are discussed. Finally, the effect of demand response in electricity prices is highlighted using a simulated case study. (author)

  8. Safety-related Innovative Nuclear Reactor Technology Elements R and D (SINTER) Network and Global HTGR R and D Network (GHTRN). Strategic benefits of international networking

    Von Lensa, W.

    1998-01-01

    Action on 'Safety-related Innovative Nuclear Reactor Technology Elements - R and D - (SINTER) Network' both aim at the identification of priority items for sustainable innovations of nuclear technologies and work-shared European collaboration structures. Such an approach can also be realised for future R and D on HTGR-related R and D under the umbrella of the IAEA as already proposed by the 'International Working Group on Gas-Cooled Reactors (IWGGCR)' in 1996 and illustrated in this paper for the construction of a 'Global HTGR R and D Network (GHTRN)'. 3 refs

  9. Final Report: Summary of Findings and Recommendations for Suction Devices for Management of Prehospital Combat Casualty Care Injuries

    2017-11-13

    Airway Final Report: Summary of Findings and Recommendations for Suction Devices for Management of Prehospital Combat Casualty Care Injuries...Consumer Style Comparison Table of Suction Pump Devices ............................. 103 Appendix H – Web Links for Images for Consumer- Style ...0022 pg. 6 Executive Summary Suction is a critical component of airway management , which is the second leading cause of preventable

  10. U.S. Annual Climatological Summaries

    National Oceanic and Atmospheric Administration, Department of Commerce — Annual Climatological Summary contains historical monthly and annual summaries for over 8000 U.S. locations. Observing stations are located in the United States of...

  11. SIMS analysis: Development and evaluation program summary

    Groenewold, G.S.; Appelhans, A.D.; Ingram, J.C.; Delmore, J.E.; Dahl, D.A.

    1996-11-01

    This report provides an overview of the ''SIMS Analysis: Development and Evaluation Program'', which was executed at the Idaho National Engineering Laboratory from mid-FY-92 to the end of FY-96. It should be noted that prior to FY-1994 the name of the program was ''In-Situ SIMS Analysis''. This report will not go into exhaustive detail regarding program accomplishments, because this information is contained in annual reports which are referenced herein. In summary, the program resulted in the design and construction of an ion trap secondary ion mass spectrometer (IT-SIMS), which is capable of the rapid analysis of environmental samples for adsorbed surface contaminants. This instrument achieves efficient secondary ion desorption by use of a molecular, massive ReO 4 - primary ion particle. The instrument manages surface charge buildup using a self-discharging principle, which is compatible with the pulsed nature of the ion trap. The instrument can achieve high selectivity and sensitivity using its selective ion storage and MS/MS capability. The instrument was used for detection of tri-n-butyl phosphate, salt cake (tank cake) characterization, and toxic metal speciation studies (specifically mercury). Technology transfer was also an important component of this program. The approach that was taken toward technology transfer was that of component transfer. This resulted in transfer of data acquisition and instrument control software in FY-94, and ongoing efforts to transfer primary ion gun and detector technology to other manufacturers

  12. Reusable Component Services

    U.S. Environmental Protection Agency — The Reusable Component Services (RCS) is a super-catalog of components, services, solutions and technologies that facilitates search, discovery and collaboration in...

  13. Safeguards Summary Event List (SSEL)

    1991-07-01

    The Safeguards Summary Event List (SSEL) provides brief summaries of several hundred safeguards-related events involving nuclear material or facilities regulated by the US Nuclear Regulatory Commission (NRC). Because of public interest, the Miscellaneous category includes a few events which involve either source material, byproduct material, or natural uranium which are exempt from safeguards requirements. Events are described under the categories of bomb-related, intrusion, missing and/or allegedly stolen, transportation, tampering/vandalism, arson, firearms, radiological sabotage, nonradiological sabotage, pre-1990 alcohol and drugs (involving reactor operators, security force members, or management persons), and miscellaneous. The information contained in the event descriptions is derived primarily from official NRC reporting channels

  14. ENDF/B summary documentation

    Kinsey, R.

    1979-07-01

    This publication provides a localized source of descriptions for the evaluations contained in the ENDF/B Library. The summary documentation presented is intended to be a more detailed description than the (File 1) comments contained in the computer readable data files, but not so detailed as the formal reports describing each ENDF/B evaluation. The summary documentations were written by the CSEWB (Cross Section Evaluation Working Group) evaluators and compiled by NNDC (National Nuclear Data Center). This edition includes documentation for materials found on ENDF/B Version V tapes 501 to 516 (General Purpose File) excluding tape 504. ENDF/B-V also includes tapes containing partial evaluations for the Special Purpose Actinide (521, 522), Dosimetry (531), Activation (532), Gas Production (533), and Fission Product (541-546) files. The materials found on these tapes are documented elsewhere. Some of the evaluation descriptions in this report contain cross sections or energy level information

  15. Visualizing Summary Statistics and Uncertainty

    Potter, K.

    2010-08-12

    The graphical depiction of uncertainty information is emerging as a problem of great importance. Scientific data sets are not considered complete without indications of error, accuracy, or levels of confidence. The visual portrayal of this information is a challenging task. This work takes inspiration from graphical data analysis to create visual representations that show not only the data value, but also important characteristics of the data including uncertainty. The canonical box plot is reexamined and a new hybrid summary plot is presented that incorporates a collection of descriptive statistics to highlight salient features of the data. Additionally, we present an extension of the summary plot to two dimensional distributions. Finally, a use-case of these new plots is presented, demonstrating their ability to present high-level overviews as well as detailed insight into the salient features of the underlying data distribution. © 2010 The Eurographics Association and Blackwell Publishing Ltd.

  16. Visualizing Summary Statistics and Uncertainty

    Potter, K.; Kniss, J.; Riesenfeld, R.; Johnson, C.R.

    2010-01-01

    The graphical depiction of uncertainty information is emerging as a problem of great importance. Scientific data sets are not considered complete without indications of error, accuracy, or levels of confidence. The visual portrayal of this information is a challenging task. This work takes inspiration from graphical data analysis to create visual representations that show not only the data value, but also important characteristics of the data including uncertainty. The canonical box plot is reexamined and a new hybrid summary plot is presented that incorporates a collection of descriptive statistics to highlight salient features of the data. Additionally, we present an extension of the summary plot to two dimensional distributions. Finally, a use-case of these new plots is presented, demonstrating their ability to present high-level overviews as well as detailed insight into the salient features of the underlying data distribution. © 2010 The Eurographics Association and Blackwell Publishing Ltd.

  17. ENDF/B summary documentation

    Kinsey, R. (comp.)

    1979-07-01

    This publication provides a localized source of descriptions for the evaluations contained in the ENDF/B Library. The summary documentation presented is intended to be a more detailed description than the (File 1) comments contained in the computer readable data files, but not so detailed as the formal reports describing each ENDF/B evaluation. The summary documentations were written by the CSEWB (Cross Section Evaluation Working Group) evaluators and compiled by NNDC (National Nuclear Data Center). This edition includes documentation for materials found on ENDF/B Version V tapes 501 to 516 (General Purpose File) excluding tape 504. ENDF/B-V also includes tapes containing partial evaluations for the Special Purpose Actinide (521, 522), Dosimetry (531), Activation (532), Gas Production (533), and Fission Product (541-546) files. The materials found on these tapes are documented elsewhere. Some of the evaluation descriptions in this report contain cross sections or energy level information. (RWR)

  18. Software component quality evaluation

    Clough, A. J.

    1991-01-01

    The paper describes a software inspection process that can be used to evaluate the quality of software components. Quality criteria, process application, independent testing of the process and proposed associated tool support are covered. Early results indicate that this technique is well suited for assessing software component quality in a standardized fashion. With automated machine assistance to facilitate both the evaluation and selection of software components, such a technique should promote effective reuse of software components.

  19. 19 CFR 210.18 - Summary determinations.

    2010-04-01

    ... 19 Customs Duties 3 2010-04-01 2010-04-01 false Summary determinations. 210.18 Section 210.18 Customs Duties UNITED STATES INTERNATIONAL TRADE COMMISSION INVESTIGATIONS OF UNFAIR PRACTICES IN IMPORT TRADE ADJUDICATION AND ENFORCEMENT Motions § 210.18 Summary determinations. (a) Motions for summary...

  20. Sexual Harassment and Organizational Outcomes Executive Summary

    2011-10-01

    quid pro quo type of Sexual harassment and Organizational, 4 sexual harassment (e.g., sexual coercion). This should drive organizational efforts to... Sexual Harassment and Organizational Outcomes Executive Summary Charlie L. Law DEFENSE EQUAL...Executive Summary] No. 99-11 Sexual harassment and Organizational, 2 Executive Summary Issue

  1. 29 CFR 1905.41 - Summary decision.

    2010-07-01

    ... OCCUPATIONAL SAFETY AND HEALTH ACT OF 1970 Summary Decisions § 1905.41 Summary decision. (a) No genuine issue... 29 Labor 5 2010-07-01 2010-07-01 false Summary decision. 1905.41 Section 1905.41 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR RULES OF...

  2. 40 CFR 1502.12 - Summary.

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Summary. 1502.12 Section 1502.12 Protection of Environment COUNCIL ON ENVIRONMENTAL QUALITY ENVIRONMENTAL IMPACT STATEMENT § 1502.12 Summary. Each environmental impact statement shall contain a summary which adequately and accurately summarizes...

  3. 40 CFR 68.155 - Executive summary.

    2010-07-01

    ... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Executive summary. 68.155 Section 68...) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.155 Executive summary. The owner or operator shall provide in the RMP an executive summary that includes a brief description of the following...

  4. 77 FR 12865 - Enforcement Actions Summary

    2012-03-02

    ...] Enforcement Actions Summary AGENCY: Transportation Security Administration, DHS. ACTION: Notice of availability. SUMMARY: The Transportation Security Administration (TSA) is providing notice that it has issued an annual summary of all enforcement actions taken by TSA under the authority granted in the...

  5. 49 CFR 1150.44 - Caption summary.

    2010-10-01

    ... 49 Transportation 8 2010-10-01 2010-10-01 false Caption summary. 1150.44 Section 1150.44... Exempt Transactions Under 49 U.S.C. 10902 for Class III Rail Carriers § 1150.44 Caption summary. The caption summary must be in the following form. The information symbolized by numbers is identified in the...

  6. 29 CFR 1904.32 - Annual summary.

    2010-07-01

    ... 29 Labor 5 2010-07-01 2010-07-01 false Annual summary. 1904.32 Section 1904.32 Labor Regulations... Requirements § 1904.32 Annual summary. (a) Basic requirement. At the end of each calendar year, you must: (1... deficiencies identified; (2) Create an annual summary of injuries and illnesses recorded on the OSHA 300 Log...

  7. 78 FR 11216 - Enforcement Actions Summary

    2013-02-15

    ...] Enforcement Actions Summary AGENCY: Transportation Security Administration, DHS. ACTION: Notice of availability. SUMMARY: The Transportation Security Administration (TSA) is providing notice that it has issued an annual summary of all enforcement actions taken by TSA under the authority granted in the...

  8. 76 FR 9357 - Enforcement Actions Summary

    2011-02-17

    ...] Enforcement Actions Summary AGENCY: Transportation Security Administration, DHS. ACTION: Notice of Availability. SUMMARY: The Transportation Security Administration (TSA) is providing notice that it has issued an annual summary of all enforcement actions taken by TSA under the authority granted in the...

  9. Regional Management Plan: Summary report

    Drobny, N.L.

    1986-01-01

    This summary report describes the results of a 16-month project to develop a Regional Management Plan for low-level radioactive waste management in a seven-state area. The seven states are Indiana, Iowa, Michigan, Minnesota, Missouri, Ohio, and Wisconsin. These states have formed the Midwest Interstate Low-Level Radioactive Waste Commission in accord with Congressional requirements established in 1980. 14 refs., 13 figs., 9 tabs

  10. ENDF/B summary documentation

    Garber, D.

    1975-10-01

    Descriptions of the evaluations contained in the ENDF/B library are given. The summary documentation presented is intended to be a more detailed description than the (File 1) comments contained in the computer-readable data files, but not so detailed as the formal reports describing each ENDF/B evaluation. The documentations were written by the CSEWG evaluators and compiled by NNCSC. Selected materials which comprise this volume include from 1 H to 244 Cm

  11. Separations innovative concepts: Project summary

    Lee, V.E. (ed.)

    1988-05-01

    This project summary includes the results of 10 innovations that were funded under the US Department's Innovative Concept Programs. The concepts address innovations that can substantially reduce the energy used in industrial separations. Each paper describes the proposed concept, and discusses the concept's potential energy savings, market applications, technical feasibility, prior work and state of the art, and future development needs.

  12. Component evaluation testing and analysis algorithms.

    Hart, Darren M.; Merchant, Bion John

    2011-10-01

    The Ground-Based Monitoring R&E Component Evaluation project performs testing on the hardware components that make up Seismic and Infrasound monitoring systems. The majority of the testing is focused on the Digital Waveform Recorder (DWR), Seismic Sensor, and Infrasound Sensor. In order to guarantee consistency, traceability, and visibility into the results of the testing process, it is necessary to document the test and analysis procedures that are in place. Other reports document the testing procedures that are in place (Kromer, 2007). This document serves to provide a comprehensive overview of the analysis and the algorithms that are applied to the Component Evaluation testing. A brief summary of each test is included to provide the context for the analysis that is to be performed.

  13. Reactor component automatic grapple

    Greenaway, P.R.

    1982-01-01

    A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment. (author)

  14. Repurposing learning object components

    Verbert, K.; Jovanovic, J.; Gasevic, D.; Duval, E.; Meersman, R.

    2005-01-01

    This paper presents an ontology-based framework for repurposing learning object components. Unlike the usual practice where learning object components are assembled manually, the proposed framework enables on-the-fly access and repurposing of learning object components. The framework supports two

  15. Compact Visualisation of Video Summaries

    Janko Ćalić

    2007-01-01

    Full Text Available This paper presents a system for compact and intuitive video summarisation aimed at both high-end professional production environments and small-screen portable devices. To represent large amounts of information in the form of a video key-frame summary, this paper studies the narrative grammar of comics, and using its universal and intuitive rules, lays out visual summaries in an efficient and user-centered way. In addition, the system exploits visual attention modelling and rapid serial visual presentation to generate highly compact summaries on mobile devices. A robust real-time algorithm for key-frame extraction is presented. The system ranks importance of key-frame sizes in the final layout by balancing the dominant visual representability and discovery of unanticipated content utilising a specific cost function and an unsupervised robust spectral clustering technique. A final layout is created using an optimisation algorithm based on dynamic programming. Algorithm efficiency and robustness are demonstrated by comparing the results with a manually labelled ground truth and with optimal panelling solutions.

  16. Summary

    Bickerton, George

    1997-01-01

    The Chernobyl accident has pushed many countries in analyzing critically their radiological emergency plans in order to identify the fields requiring amelioration or reinforcement. A common topic was the necessity of informing and drilling efficiently the civil population as well as different intervening agents against effects of nuclear accidents. It was stressed that the health and safety of populations, potentially most exposed, constitute a top priority, followed by the protection of food storage. The majority of the countries, were the management of public affairs is effected at two or more echelons, is confronted with the difficulty of developing plans clearly defining the missions and responsibilities of different administration levels as well as the interactions between them in case of emergency. Emphasized are also the requirements of information tools able of computing the contamination of foods or forages taking into account of factors like annual season, the phase of crop growth and the meteorological conditions. Obviously, such programs permit forecasting, surveying and evaluating the contamination and consequently, establishing the proper level of intervention. Also, the question of intervention thresholds was approached as well as the necessity of harmonizing intervention on international scale. A number of topics were mentioned to be under current study among which the relations between soil contamination and radionuclide concentration in milk and forage as well as the methods of managing the highly cesium-contaminated milk. Finally, it was argued for the necessity of ensuring the population confidence in the measures of intervention as well as in the indications of competent officials in charge with emergency actions

  17. Summary

    2002-01-01

    This report summarizes the progress made that resulted in the improvement of the organizational and analytical performance on Nuclear Analytical Laboratories at the Ankara Nuclear Research and Training Centre in accordance with ISO 17025 requirements. This report lists the main accomplishments and presents future plans of the Laboratory

  18. Summary

    In explaining this, we determine spin (frequency) of black holes which is a parameter of our model. For the best observational fit, we obtain the spin parameter of black holes ranges: 0.6 ≤ a ≤ 0.8: None of them corresponds to an extremally (a~1) rotating black hole. This is a test of Einstein's theory of General Relativity.

  19. SUMMARY

    Health services, whether preventive, curative or rehabilitative, ... price (premium) that they will be prepared to pay for health ... nurses, 4% each accountants, administrators, research ..... Ghana. Exchange Health Economics and ... approaches.

  20. 31 CFR 26.3 - Availability of Environmental Impact Assessment Summaries (EIA Summaries) and Environmental...

    2010-07-01

    ... 31 Money and Finance: Treasury 1 2010-07-01 2010-07-01 false Availability of Environmental Impact Assessment Summaries (EIA Summaries) and Environmental Impact Assessments (EIAs). 26.3 Section 26.3 Money and... DEVELOPMENT BANDS (MDBs) § 26.3 Availability of Environmental Impact Assessment Summaries (EIA Summaries) and...

  1. Significance of BETA and GAMMA dose on environmental qualification of components

    Aydogdu, K.M.; Tsang, K.T.

    1999-01-01

    Safety-related systems and components that are required to perform safety functions during accident conditions must be designed to withstand the harsh environmental conditions that occur as a consequence of the accident. Where these conditions are 'harsh', and equipment operability can potentially be affected by the post-accident environment environmental qualification of the equipment must be conducted to demonstrate that the required safety function can be maintained. It is also understood that non-safety related equipment that affects, or prevents, the satisfactory operation of a safety-related system should also withstand the 'harsh' environmental conditions caused by an appropriate design-basis accident. There are essentially two types of requirements that must be satisfied to qualify equipment or components to withstand radiation damage, namely economic requirements and safety requirements. The general objective of the economic requirement is to reduce maintenance cost and to maximize component life during reactor operation. The general objective of the safety requirement is that the equipment should be qualified to withstand the harsh post-accident environmental conditions and should function properly for the appropriate length of time after a design-basis accident has occurred. To address the economic factors - i.e., to reduce maintenance costs and to maximize component life - the radiation dose rates to equipment are calculated throughout the reactor building and the service building during reactor operation. These are also used for the safety requirement purpose, to assess radiation ageing of safety-related components caused by degradation of material properties with time at radiation exposure. To address the safety requirement, the dose-rate estimates and accumulated doses after a LOCA coincident with loss-of-emergency-core cooling (LOECC) are provided. The harsh post-accident environmental conditions defined for environmental qualification of components

  2. Physical characteristics of non-fuel assembly reactor components

    Hawkes, E.C.

    1994-09-01

    The primary objective of this report is to enhance the utility of the Characteristics Data Base (CDB). This has been accomplished by providing a pictorial representation of the principal non-fuel assembly (NFA) components along with a tabular summary of key information about each type of component. This report is intended for use as an adjunct to the CDB. Toward this end, the report may be used either as a complement to the detailed descriptions in the CDB, or as a stand-alone document that acts as an illustrated abstract of the CDB. Line drawings of major NFA components are included. Data not provided in the CDB are also included. Summary descriptions of each component are given in tabular format

  3. Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-Category I systems for Turkey Point Nuclear Power Plant, Units 3 and 4

    Collins, E.K.

    1979-08-01

    Three separate reviews of the Turkey Point Units 3 and 4 were conducted by the FPLCO between 1972 and 1975. Initially, at the request of NBC in 1972, the FPLCO reviewed several water systems as sources of flooding. Subsequently, as a result of an abnormal occurrence, the drainage system was reviewed. Finally, the facilities were again reviewed at NRC's request and both the potential sources of flooding and safety-related equipment which could be damaged by flooding were identified. The sources of flooding and the appropriate safety equipment are discussed. An evaluation is presented of measures that were taken by FPLCO to minimize the danger of flooding and to protect safety-related equipment

  4. Why Summary Comorbidity Measures Such As the Charlson Comorbidity Index and Elixhauser Score Work.

    Austin, Steven R; Wong, Yu-Ning; Uzzo, Robert G; Beck, J Robert; Egleston, Brian L

    2015-09-01

    Comorbidity adjustment is an important component of health services research and clinical prognosis. When adjusting for comorbidities in statistical models, researchers can include comorbidities individually or through the use of summary measures such as the Charlson Comorbidity Index or Elixhauser score. We examined the conditions under which individual versus summary measures are most appropriate. We provide an analytic proof of the utility of comorbidity summary measures when used in place of individual comorbidities. We compared the use of the Charlson and Elixhauser scores versus individual comorbidities in prognostic models using a SEER-Medicare data example. We examined the ability of summary comorbidity measures to adjust for confounding using simulations. We devised a mathematical proof that found that the comorbidity summary measures are appropriate prognostic or adjustment mechanisms in survival analyses. Once one knows the comorbidity score, no other information about the comorbidity variables used to create the score is generally needed. Our data example and simulations largely confirmed this finding. Summary comorbidity measures, such as the Charlson Comorbidity Index and Elixhauser scores, are commonly used for clinical prognosis and comorbidity adjustment. We have provided a theoretical justification that validates the use of such scores under many conditions. Our simulations generally confirm the utility of the summary comorbidity measures as substitutes for use of the individual comorbidity variables in health services research. One caveat is that a summary measure may only be as good as the variables used to create it.

  5. Contribution to a quantitative assessment model for reliability-based metrics of electronic and programmable safety-related functions; Contribution a un modele d'evaluation quantitative des performances fiabilistes de fonctions electroniques et programmables dediees a la securite

    Hamidi, K

    2005-10-15

    The use of fault-tolerant EP architectures has induced growing constraints, whose influence on reliability-based performance metrics is no more negligible. To face up the growing influence of simultaneous failure, this thesis proposes, for safety-related functions, a new-trend assessment method of reliability, based on a better taking into account of time-aspect. This report introduces the concept of information and uses it to interpret the failure modes of safety-related function as the direct result of the initiation and propagation of erroneous information until the actuator-level. The main idea is to distinguish the apparition and disappearance of erroneous states, which could be defined as intrinsically dependent of HW-characteristic and maintenance policies, and their possible activation, constrained through architectural choices, leading to the failure of safety-related function. This approach is based on a low level on deterministic SED models of the architecture and use non homogeneous Markov chains to depict the time-evolution of probabilities of errors. (author)

  6. Contribution to a quantitative assessment model for reliability-based metrics of electronic and programmable safety-related functions; Contribution a un modele d'evaluation quantitative des performances fiabilistes de fonctions electroniques et programmables dediees a la securite

    Hamidi, K

    2005-10-15

    The use of fault-tolerant EP architectures has induced growing constraints, whose influence on reliability-based performance metrics is no more negligible. To face up the growing influence of simultaneous failure, this thesis proposes, for safety-related functions, a new-trend assessment method of reliability, based on a better taking into account of time-aspect. This report introduces the concept of information and uses it to interpret the failure modes of safety-related function as the direct result of the initiation and propagation of erroneous information until the actuator-level. The main idea is to distinguish the apparition and disappearance of erroneous states, which could be defined as intrinsically dependent of HW-characteristic and maintenance policies, and their possible activation, constrained through architectural choices, leading to the failure of safety-related function. This approach is based on a low level on deterministic SED models of the architecture and use non homogeneous Markov chains to depict the time-evolution of probabilities of errors. (author)

  7. Plutonium focus area: Technology summary

    1996-03-01

    To ensure research and development programs focus on the most pressing environmental restoration and waste management problems at the U.S. Department of Energy (DOE), the Assistant Secretary for the Office of Environmental Management (EM) established a working group in August 1993 to implement a new approach to research and technology development. As part of this approach, EM developed a management structure and principles that led to creation of specific focus areas. These organizations were designed to focus scientific and technical talent throughout DOE and the national scientific community on major environmental restoration and waste management problems facing DOE. The focus area approach provides the framework for inter-site cooperation and leveraging of resources on common problems. After the original establishment of five major focus areas within the Office of Technology Development (EM-50), the Nuclear Materials Stabilization Task Group (NMSTG, EM-66) followed EM-50's structure and chartered the Plutonium Focus Area (PFA). NMSTG's charter to the PFA, described in detail later in this book, plays a major role in meeting the EM-66 commitments to the Defense Nuclear Facilities Safety Board (DNFSB). The PFA is a new program for FY96 and as such, the primary focus of revision 0 of this Technology Summary is an introduction to the Focus Area; its history, development, and management structure, including summaries of selected technologies being developed. Revision 1 to the Plutonium Focus Area Technology Summary is slated to include details on all technologies being developed, and is currently planned for release in August 1996. The following report outlines the scope and mission of the Office of Environmental Management, EM-60, and EM-66 organizations as related to the PFA organizational structure

  8. Fusion Plasma Theory project summaries

    1993-10-01

    This Project Summary book is a published compilation consisting of short descriptions of each project supported by the Fusion Plasma Theory and Computing Group of the Advanced Physics and Technology Division of the Department of Energy, Office of Fusion Energy. The summaries contained in this volume were written by the individual contractors with minimal editing by the Office of Fusion Energy. Previous summaries were published in February of 1982 and December of 1987. The Plasma Theory program is responsible for the development of concepts and models that describe and predict the behavior of a magnetically confined plasma. Emphasis is given to the modelling and understanding of the processes controlling transport of energy and particles in a toroidal plasma and supporting the design of the International Thermonuclear Experimental Reactor (ITER). A tokamak transport initiative was begun in 1989 to improve understanding of how energy and particles are lost from the plasma by mechanisms that transport them across field lines. The Plasma Theory program has actively-participated in this initiative. Recently, increased attention has been given to issues of importance to the proposed Tokamak Physics Experiment (TPX). Particular attention has been paid to containment and thermalization of fast alpha particles produced in a burning fusion plasma as well as control of sawteeth, current drive, impurity control, and design of improved auxiliary heating. In addition, general models of plasma behavior are developed from physics features common to different confinement geometries. This work uses both analytical and numerical techniques. The Fusion Theory program supports research projects at US government laboratories, universities and industrial contractors. Its support of theoretical work at universities contributes to the office of Fusion Energy mission of training scientific manpower for the US Fusion Energy Program.

  9. Fusion Plasma Theory project summaries

    1993-10-01

    This Project Summary book is a published compilation consisting of short descriptions of each project supported by the Fusion Plasma Theory and Computing Group of the Advanced Physics and Technology Division of the Department of Energy, Office of Fusion Energy. The summaries contained in this volume were written by the individual contractors with minimal editing by the Office of Fusion Energy. Previous summaries were published in February of 1982 and December of 1987. The Plasma Theory program is responsible for the development of concepts and models that describe and predict the behavior of a magnetically confined plasma. Emphasis is given to the modelling and understanding of the processes controlling transport of energy and particles in a toroidal plasma and supporting the design of the International Thermonuclear Experimental Reactor (ITER). A tokamak transport initiative was begun in 1989 to improve understanding of how energy and particles are lost from the plasma by mechanisms that transport them across field lines. The Plasma Theory program has actively-participated in this initiative. Recently, increased attention has been given to issues of importance to the proposed Tokamak Physics Experiment (TPX). Particular attention has been paid to containment and thermalization of fast alpha particles produced in a burning fusion plasma as well as control of sawteeth, current drive, impurity control, and design of improved auxiliary heating. In addition, general models of plasma behavior are developed from physics features common to different confinement geometries. This work uses both analytical and numerical techniques. The Fusion Theory program supports research projects at US government laboratories, universities and industrial contractors. Its support of theoretical work at universities contributes to the office of Fusion Energy mission of training scientific manpower for the US Fusion Energy Program

  10. Fusion plasma theory project summaries

    1993-10-01

    This Project Summary book is a published compilation consisting of short descriptions of each project supported by the Fusion Plasma Theory and Computing Group of the Advanced Physics and Technology Division of the Department of Energy, Office of Fusion Energy. The summaries contained in this volume were written by the individual contractors with minimal editing by the Office of Fusion Energy. Previous summaries were published in February of 1982 and December of 1987. The Plasma Theory program is responsible for the development of concepts and models that describe and predict the behavior of a magnetically confined plasma. Emphasis is given to the modelling and understanding of the processes controlling transport of energy and particles in a toroidal plasma and supporting the design of the International Thermonuclear Experimental Reactor (ITER). A tokamak transport initiative was begun in 1989 to improve understanding of how energy and particles are lost from the plasma by mechanisms that transport them across field lines. The Plasma Theory program has actively participated in this initiative. Recently, increased attention has been given to issues of importance to the proposed Tokamak Physics Experiment (TPX). Particular attention has been paid to containment and thermalization of fast alpha particles produced in a burning fusion plasma as well as control of sawteeth, current drive, impurity control, and design of improved auxiliary heating. In addition, general models of plasma behavior are developed from physics features common to different confinement geometries. This work uses both analytical and numerical techniques. The Fusion Theory program supports research projects at U.S. government laboratories, universities and industrial contractors. Its support of theoretical work at universities contributes to the office of Fusion Energy mission of training scientific manpower for the U.S. Fusion Energy Program.

  11. Supply chain components

    Vieraşu, T.; Bălăşescu, M.

    2011-01-01

    In this article I will go through three main logistics components, which are represented by: transportation, inventory and facilities, and the three secondary logistical components: information, production location, price and how they determine performance of any supply chain. I will discuss then how these components are used in the design, planning and operation of a supply chain. I will also talk about some obstacles a supply chain manager may encounter.

  12. Supply chain components

    Vieraşu, T.

    2011-01-01

    Full Text Available In this article I will go through three main logistics components, which are represented by: transportation, inventory and facilities, and the three secondary logistical components: information, production location, price and how they determine performance of any supply chain. I will discuss then how these components are used in the design, planning and operation of a supply chain. I will also talk about some obstacles a supply chain manager may encounter.

  13. Control component retainer

    Walton, L.A.; King, R.A.

    1983-01-01

    An apparatus is described for retaining an undriven control component assembly disposed in a fuel assembly in a nuclear reactor of the type having a core grid plate. The first part of the mechanism involves a housing for the control component and the second part is a brace with a number of arms that reach under the grid plate. The brace and the housing are coupled together to firmly hold the control components in place even under strong flows of th coolant

  14. Experimental plasma research project summaries

    1982-10-01

    The experimental plasma Research Branch has responsibility for developing a broad range of experimental data and new experimental techniques that are required for operating and interpreting present large-scale confinement experiments, and for designing future deuterium-tritium burining facilities. The Branch pursued these objectives by supporting research in DOE laboratories, other Federal laboratories, other Federal laboratories, universities, and private industry. Initiation and renewal of research projects are primarily through submission of unsolicited proposals by these institutions to DOE. Summaries of these projects are given

  15. Summary of the Accelerator Meeting

    Lia Merminga

    2006-01-01

    The summary of the paper is: (1) A compelling scientific case is developing for a high luminosity, polarized electron-ion collider, to address fundamental questions in hadron Physics. (2) Much progress over the past years: Design concepts are maturing through innovation and design optimization. (3) Electron cooling is prerequisite for all EIC design scenarios. A rigorous electron cooling R and D program established at BNL. (4) Important to continue collaboration and cross-fertilization of ideas among different design options. (5) Thank you to all the speakers for outstanding and thought-provoking presentations

  16. Product Operations Status Summary Metrics

    Takagi, Atsuya; Toole, Nicholas

    2010-01-01

    The Product Operations Status Summary Metrics (POSSUM) computer program provides a readable view into the state of the Phoenix Operations Product Generation Subsystem (OPGS) data pipeline. POSSUM provides a user interface that can search the data store, collect product metadata, and display the results in an easily-readable layout. It was designed with flexibility in mind for support in future missions. Flexibility over various data store hierarchies is provided through the disk-searching facilities of Marsviewer. This is a proven program that has been in operational use since the first day of the Phoenix mission.

  17. Advanced fusion concepts: project summaries

    1980-12-01

    This report contains descriptions of the activities of all the projects supported by the Advanced Fusion Concepts Branch of the Office of Fusion Energy, US Department of Energy. These descriptions are project summaries of each of the individual projects, and contain the following: title, principle investigators, funding levels, purpose, approach, progress, plans, milestones, graduate students, graduates, other professional staff, and recent publications. Information is given for each of the following programs: (1) reverse-field pinch, (2) compact toroid, (3) alternate fuel/multipoles, (4) stellarator/torsatron, (5) linear magnetic fusion, (6) liners, and (7) Tormac

  18. Summary inside IBM's historic turnaround

    2014-01-01

    This work offers a summary of the book "WHO SAYS ELEPHANTS CAN'T DANCE? Inside IBM's Historic Turnaround" by Louis Gerstner.In nine years as the chairman and CEO of International Business Machine Corporation (IBM), Louis Gerstner brought about a dramatic change in the company's fortunes. When he took charge, IBM was on the verge of extinction as the victim of rapid changes in the computer industry. However, instead of breaking up IBM as most analysts were suggesting, Gerstner and his management team turned the company around and restored it to a position of power and influence within the indu

  19. National stakeholder workshop summary report

    NONE

    1996-06-01

    This is a summary of the plenary sessions and small group discussion sessions from the fourth National Stakeholder Workshop sponsored by the DOE Office of Worker and Community Transition held in Atlanta, Georgia on March 13--15, 1996. Topics of the sessions included work force planning and restructuring, worker participation in health and safety, review of actions and commitments, lessons learned in collective bargaining agreements, work force restructuring guidance, work force planning, update on community transition activities. Also included are appendices listing the participants and DOE contacts.

  20. Component design for LMFBR's

    Fillnow, R.H.; France, L.L.; Zerinvary, M.C.; Fox, R.O.

    1975-01-01

    Just as FFTF has prototype components to confirm their design, FFTF is serving as a prototype for the design of the commercial LMFBR's. Design and manufacture of critical components for the FFTF system have been accomplished primarily using vendors with little or no previous experience in supplying components for high temperature sodium systems. The exposure of these suppliers, and through them a multitude of subcontractors, to the requirements of this program has been a necessary and significant step in preparing American industry for the task of supplying the large mechanical components required for commercial LMFBR's

  1. Hot gas path component

    Lacy, Benjamin Paul; Kottilingam, Srikanth Chandrudu; Porter, Christopher Donald; Schick, David Edward

    2017-09-12

    Various embodiments of the disclosure include a turbomachine component. and methods of forming such a component. Some embodiments include a turbomachine component including: a first portion including at least one of a stainless steel or an alloy steel; and a second portion joined with the first portion, the second portion including a nickel alloy including an arced cooling feature extending therethrough, the second portion having a thermal expansion coefficient substantially similar to a thermal expansion coefficient of the first portion, wherein the arced cooling feature is located within the second portion to direct a portion of a coolant to a leakage area of the turbomachine component.

  2. Evaluation of the national component of the climate protection initiative of the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety. Summary. Final Report 2012; Evaluierung des nationalen Teils der Klimaschutzinitiative des Bundesministeriums fuer Umwelt, Naturschutz und Reaktorsicherheit. Zusammenfassung. Endbericht 2012

    Schumacher, Katja; Repenning, Julia; Matthes, Felix C. [Oeko-Institut - Institut fuer angewandte Oekologie, Berlin (Germany)] [and others

    2012-10-19

    The National Climate Initiative of Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (Berlin, Federal Republic of Germany) is an important component of the integrated energy and climate protection programme of the Federal Government as well as the energy concept of the Federal Government or the energy policy turnaround, respectively. The National Climate Initiative has to reduce the German greenhouse gas emissions to 40 % in the year 2020 in comparison to the year 1990. The evaluation of the National Climate Initiative refers to 21 single projects, 4 regulations and 4 accumulations of capital in the support years 2008 to 2011. 900 million Euro of federal funds were invested in these measures. The evaluated portfolio is the first generation of the National Climate Initiatives. Since then, the concept of the National Climate Initiative was refined. In the course of the implementation and evaluation of these first activities important experiences were gathered yielded in a further design of the National Climate Initiative.

  3. Plutonium focus area. Technology summary

    1997-09-01

    The Assistant Secretary for the Office of Environmental Management (EM) at the U.S. Department of Energy (DOE) chartered the Plutonium Focus Area (PFA) in October 1995. The PFA open-quotes...provides for peer and technical reviews of research and development in plutonium stabilization activities...close quotes In addition, the PFA identifies and develops relevant research and technology. The purpose of this document is to focus attention on the requirements used to develop research and technology for stabilization, storage, and preparation for disposition of nuclear materials. The PFA Technology Summary presents the approach the PFA uses to identify, recommend, and review research. It lists research requirements, research being conducted, and gaps where research is needed. It also summarizes research performed by the PFA in the traditional research summary format. This document encourages researchers and commercial enterprises to do business with PFA by submitting research proposals or open-quotes white papers.close quotes In addition, it suggests ways to increase the likelihood that PFA will recommend proposed research to the Nuclear Materials Stabilization Task Group (NMSTG) of DOE

  4. Components of Sexual Identity

    Shively, Michael G.; DeCecco, John P.

    1977-01-01

    This paper examines the four components of sexual identity: biological sex, gender identity, social sex-role, and sexual orientation. Theories about the development of each component and how they combine and conflict to form the individual's sexual identity are discussed. (Author)

  5. Towards Cognitive Component Analysis

    Hansen, Lars Kai; Ahrendt, Peter; Larsen, Jan

    2005-01-01

    Cognitive component analysis (COCA) is here defined as the process of unsupervised grouping of data such that the ensuing group structure is well-aligned with that resulting from human cognitive activity. We have earlier demonstrated that independent components analysis is relevant for representing...

  6. Z-1 Prototype Space Suit Testing Summary

    Ross, Amy

    2013-01-01

    The Advanced Space Suit team of the NASA-Johnson Space Center performed a series of test with the Z-1 prototype space suit in 2012. This paper discusses, at a summary level, the tests performed and results from those tests. The purpose of the tests were two-fold: 1) characterize the suit performance so that the data could be used in the downselection of components for the Z-2 Space Suit and 2) develop interfaces with the suitport and exploration vehicles through pressurized suit evaluations. Tests performed included isolated and functional range of motion data capture, Z-1 waist and hip testing, joint torque testing, CO2 washout testing, fit checks and subject familiarizations, an exploration vehicle aft deck and suitport controls interface evaluation, delta pressure suitport tests including pressurized suit don and doff, and gross mobility and suitport ingress and egress demonstrations in reduced gravity. Lessons learned specific to the Z-1 prototype and to suit testing techniques will be presented.

  7. Summary pamphlet: 1995 site environmental report

    1996-01-01

    As required in U.S. Department of Energy (DOE) Order 5400.1, an Annual Site Environmental Report (ASER) has been prepared for Sandia National Laboratories/New Mexico (SNL/NM) for 1995. The ASER represents a key component of the DOE's effort to keep the public informed about environmental efforts and compliance status at SNL/NM. This booklett was prepared by the Environmental Operations Center of SNL/NM and reviewed by Community Relations and Risk Management. Suggestions were incorporated from the students of New Futures High School as a part of the Environmental Education Program. This work is supported by the DOE under Contract DE-AC04-94AL85000. A copy of the ASER can be obtained by calling the Environmental Monitoring and Reporting Department at 848-0927. This pamphlet provides a brief summary of the 1995 SNL/NM environmental programs and monitoring results. Additional copies of this pamphlet may be obtained by calling the number above

  8. Traffic management simulation development : summary.

    2011-01-01

    Increasingly, Florida traffic is monitored electronically by components of the Intelligent Traffic System (ITS), which send data to regional traffic management centers and assist management of traffic flows and incident response using software called...

  9. Specialists’ Meeting on Demonstration of Structural Integrity under Normal and Faulted Conditions. Summary Report

    1981-03-01

    The Specialists' Meeting on ''Demonstration of Structural Integrity under Normal and Faulted Conditions'' was held at Chester, United Kingdom on 3-5 June 1980. The meeting was sponsored by the International Atomic Energy Agency (IAEA) on the recommendation of the International Working Group on Past Reactors (IWGFR). Twenty-one participants from France, the Federal Republic of Germany, Italy, Japan, the Netherlands, the United Kingdom, the United States of America and two international organizations, CEC and IAEA, attended. The purpose of the meeting was to review and discuss methods for assessing the integrity of the LMFBR safety-related structures during normal and abnormal operation, especially in the presence of defects, and to recommend future development. The technical sessions were divided into four topical sessions as follows: 1. National Review Presentations on Demonstration of Structural Integrity; 2. Material Properties; 3. Structural Analysis; 4. Design Approaches and Assessment Experience. During the meeting papers were presented by the participants on behalf of their countries or organizations. Each presentation was followed by an open discussion in the subject covered by the paper and subsequently, session summaries were drafted. After the formal sessions were completed, a final discussion session was held and general conclusions and recommendations were reached by consensus. Session summaries, general conclusions and recommendations, national review papers presented during the first session as well as the agenda of the meeting and the list of participants are given

  10. Components and renewal parts in the nuclear power industry

    Clark, T.F. Jr.

    1986-01-01

    This paper indicates that the nuclear parts industry has been forced to make major investments in time, personnel and financial resources in order to solve short term/emergency procurement problems. What is required, as was previously indicated, is a coordinated industry-wide effort toward long range planning and implementation of a program that addresses these issues. The industry is developing programs directed toward inventory optimization and ''innovative-creative'' financing of manufacturing inventory/work-in-process in an effort to significantly reduce delivery lead times. Product transition, utilization of cancelled plant equipment, equipment qualification programs, and dedication of commercially manufactured/procured parts and components for safety related application continue to be major elements of our program to support current utility requirements

  11. GCS component development cycle

    Rodríguez, Jose A.; Macias, Rosa; Molgo, Jordi; Guerra, Dailos; Pi, Marti

    2012-09-01

    The GTC1 is an optical-infrared 10-meter segmented mirror telescope at the ORM observatory in Canary Islands (Spain). First light was at 13/07/2007 and since them it is in the operation phase. The GTC control system (GCS) is a distributed object & component oriented system based on RT-CORBA8 and it is responsible for the management and operation of the telescope, including its instrumentation. GCS has used the Rational Unified process (RUP9) in its development. RUP is an iterative software development process framework. After analysing (use cases) and designing (UML10) any of GCS subsystems, an initial component description of its interface is obtained and from that information a component specification is written. In order to improve the code productivity, GCS has adopted the code generation to transform this component specification into the skeleton of component classes based on a software framework, called Device Component Framework. Using the GCS development tools, based on javadoc and gcc, in only one step, the component is generated, compiled and deployed to be tested for the first time through our GUI inspector. The main advantages of this approach are the following: It reduces the learning curve of new developers and the development error rate, allows a systematic use of design patterns in the development and software reuse, speeds up the deliverables of the software product and massively increase the timescale, design consistency and design quality, and eliminates the future refactoring process required for the code.

  12. 2-component heating systems

    Radtke, W

    1987-03-01

    The knowledge accumulated only recently of the damage to buildings and the hazards of formaldehyde, radon and hydrocarbons has been inducing louder calls for ventilation, which, on their part, account for the fact that increasing importance is being attached to the controlled ventilation of buildings. Two-component heating systems provide for fresh air and thermal comfort in one. While the first component uses fresh air blown directly and controllably into the rooms, the second component is similar to the Roman hypocaustic heating systems, meaning that heated outer air is circulating under the floor, thus providing for hot surfaces and thermal comfort. Details concerning the two-component heating system are presented along with systems diagrams, diagrams of the heating system and tables identifying the respective costs. Descriptions are given of the two systems components, the fast heat-up, the two-component made, the change of air, heat recovery and control systems. Comparative evaluations determine the differences between two-component heating systems and other heating systems. Conclusive remarks are dedicated to energy conservation and comparative evaluations of costs. (HWJ).

  13. Advanced Fusion Concepts project summaries. FY 1983

    1983-06-01

    This report contains descriptions of the activities of all the projects supported by the Advanced Fusion Concepts Branch of the Office of Fusion Energy, US Department of Energy. These descriptions are project summaries of each of the individual projects, and contain the following: title, principle investigators, funding levels, purpose, approach, progress, plans, milestones, graduate studients, graduates, other professional staff, and recent publications. The individual project summaries are prepared by the principle investigators in collaboration with the Advanced Fusion Concepts (AFC) Branch. In addition to the project summaries, statements of branch objectives, and budget summaries are also provided

  14. Summary of longitudinal instabilities workshop

    Chasman, R.

    1976-01-01

    A five-day ISABELLE workshop on longitudinal instabilities was held at Brookhaven, August 9-13, 1976. About a dozen outside accelerator experts, both from Europe and the U.S.A., joined the local staff for discussions of longitudinal instabilities in ISABELLE. An agenda of talks was scheduled for the first day of the workshop. Later during the week, a presentation was given on the subject ''A more rigorous treatment of Landau damping in longitudinal beam instabilities''. A few progress meetings were held in which disagreements regarding calculations of coupling impedances were clarified. A summary session was held on the last day. Heavy emphasis was put on single bunched beam instabilities in the microwave region extending above the cut-off frequency of the ISABELLE vacuum chamber.

  15. CNA/CNS conference summaries

    1992-01-01

    This volume contains summaries of papers presented at the 32. annual conference of the Canadian Nuclear Association and the 13. annual conference of the Canadian Nuclear Society. The full proceedings, and the individual papers contained therein, have been abstracted separately. Sessions on the following topics are included: Plenary; The international CANDU program; Canadian used fuel management program; Public information advocates; Fuel and electricity supply; In which direction should reactors advance?; Canadian advanced nuclear research program; International cooperation in operations; Safety in design, operation, regulation; Renovation of operating stations; Reactor physics; New concepts and Technology; Fuel behaviour; Reactor design; Safety analysis; Fuel channel behaviour; Equipment and design qualification; Compliance and licensing; Fusion science and technology; Darlington assessment; Plant aging and life assessment; Thermalhydraulic modelling and analysis; Diagnostics and data management; Operator training and certification

  16. Plasma and neutralization effects: summary

    Tidman, D.A.

    1978-01-01

    The plasma working group considered the question of whether an intense heavy ion beam could be transported and accurately focussed across a target chamber radius of approximately 10 m on to a pellet of radius approximately 0.1 cm at the center of the chamber (a typical beam was taken as 3 kA, 40 GeV uranium injected into the reactor vessel with initial beam radius approximately 10 cm). Here we give a brief summary of our considerations. The conclusions were that focussing through relatively dense reactor chamber gases appears to be possible. Instabilities, if they arise, are expected only within the last few 10's of cm from the pellet, by which time they are unlikely to significantly degrade the beam focussing

  17. Summary of presentations and discussions

    Takeuchi, Mitsuo

    2008-01-01

    In December 2007, the Forum on Stakeholder Confidence discussed its theme entitled 'Link between research, development and demonstration (RD and D) and stakeholder confidence'. It was remarked that regulators need a technical demonstration to aid in evaluating the safety case. Local stakeholders appreciate the opportunity to visualise technological arrangements. In both cases, demonstration adds to confidence in the feasibility of solutions. Some believe there is an important role for analogues in communication with stakeholders, if handled with integrity. To explore and benchmark current practices, it was decided to hold a topical session at the 9. regular meeting of the FSC on 4 June 2008 regarding the use of analogues for confidence building. The session opened with an introductory presentation by the session rapporteur. This incorporated input provided for the purpose by FSC members in cooperation with their country's representative to the NEA RWMC 'Integration Group on the Safety Case'. Three speakers then presented the various uses of analogues by implementers, regulators and scientists to build their own confidence; a fourth speaker dealt with the experience of using natural analogues in public information. The presentations addressed the use of analogues in the field of geological disposal of high-level waste (HLW) and long-lived intermediate level (ILW-LL) radioactive waste. Then the FSC participants split into two working groups for discussion. The outcome of these discussions was reported in plenary on 6 June 2008 and it was agreed to publish proceedings of the session. The present summary, prepared by the session rapporteur with input from the NEA Secretariat, captures the main points heard in the course of the event. It combines data from the formal presentations and remarks made in discussion. The latter represent viewpoints expressed by a group whose primary focus is not natural analogues but rather stakeholder interests. The summary and viewpoints

  18. Experimental plasma research project summaries

    1992-06-01

    This is the latest in a series of Project Summary books going back to 1976 and is the first after a hiatus of several years. They are published to provide a short description of each project supported by the Experimental Plasma Research Branch of the Division of Applied Plasma Physics in the Office of Fusion Energy. The Experimental Plasma Research Branch seeks to provide a broad range of experimental data, physics understanding, and new experimental techniques that contribute to operation, interpretation, and improvement of high temperature plasma as a source of fusion energy. In pursuit of these objectives, the branch supports research at universities, DOE laboratories, other federal laboratories and industry. About 70 percent of the funds expended are spent at universities and a significant function of this program is the training of students in fusion physics. The branch supports small- and medium-scale experimental studies directly related to specific critical plasma issues of the magnetic fusion program. Plasma physics experiments are conducted on transport of particles and energy within plasma and innovative approaches for operating, controlling, and heating plasma are evaluated for application to the larger confinement devices of the magnetic fusion program. New diagnostic approaches to measuring the properties of high temperature plasmas are developed to the point where they can be applied with confidence on the large-scale confinement experiments. Atomic data necessary for impurity control, interpretation of diagnostic data, development of heating devices, and analysis of cooling by impurity ion radiation are obtained. The project summaries are grouped into these three categories of plasma physics, diagnostic development and atomic physics

  19. Replaceable LMFBR core components

    Evans, E.A.; Cunningham, G.W.

    1976-01-01

    Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

  20. Explosive Components Facility

    Federal Laboratory Consortium — The 98,000 square foot Explosive Components Facility (ECF) is a state-of-the-art facility that provides a full-range of chemical, material, and performance analysis...

  1. Probability of inadvertent operation of electrical components in harsh environments

    Knoll, A.

    1989-01-01

    Harsh environment, which means humidity and high temperature, may and will affect unsealed electrical components by causing leakage ground currents in ungrounded direct current systems. The concern in a nuclear power plant is that such harsh environment conditions could cause inadvertent operation of normally deenergized components, which may have a safety-related isolation function. Harsh environment is a common cause failure, and one way to approach the problem is to assume that all the unsealed electrical components will simultaneously and inadvertently energize as a result of the environmental common cause failure. This assumption is unrealistically conservative. Test results indicated that insulating resistences of any terminal block in harsh environments have a random distribution in the range of 1 to 270 kΩ, with a mean value ∼59 kΩ. The objective of this paper is to evaluate a realistic conditional failure probability for inadvertent operation of electrical components in harsh environments. This value will be used thereafter in probabilistic safety evaluations of harsh environment events and will replace both the overconservative common cause probability of 1 and the random failure probability used for mild environments

  2. Component fragility research program

    Tsai, N.C.; Mochizuki, G.L.; Holman, G.S.

    1989-11-01

    To demonstrate how ''high-level'' qualification test data can be used to estimate the ultimate seismic capacity of nuclear power plant equipment, we assessed in detail various electrical components tested by the Pacific Gas ampersand Electric Company for its Diablo Canyon plant. As part of our Phase I Component Fragility Research Program, we evaluated seismic fragility for five Diablo Canyon components: medium-voltage (4kV) switchgear; safeguard relay board; emergency light battery pack; potential transformer; and station battery and racks. This report discusses our Phase II fragility evaluation of a single Westinghouse Type W motor control center column, a fan cooler motor controller, and three local starters at the Diablo Canyon nuclear power plant. These components were seismically qualified by means of biaxial random motion tests on a shaker table, and the test response spectra formed the basis for the estimate of the seismic capacity of the components. The seismic capacity of each component is referenced to the zero period acceleration (ZPA) and, in our Phase II study only, to the average spectral acceleration (ASA) of the motion at its base. For the motor control center, the seismic capacity was compared to the capacity of a Westinghouse Five-Star MCC subjected to actual fragility tests by LLNL during the Phase I Component Fragility Research Program, and to generic capacities developed by the Brookhaven National Laboratory for motor control center. Except for the medium-voltage switchgear, all of the components considered in both our Phase I and Phase II evaluations were qualified in their standard commercial configurations or with only relatively minor modifications such as top bracing of cabinets. 8 refs., 67 figs., 7 tabs

  3. Refractory alloy component fabrication

    Young, W.R.

    1984-01-01

    Purpose of this report is to describe joining procedures, primarily welding techniques, which were developed to construct reliable refractory alloy components and systems for advanced space power systems. Two systems, the Nb-1Zr Brayton Cycle Heat Receiver and the T-111 Alloy Potassium Boiler Development Program, are used to illustrate typical systems and components. Particular emphasis is given to specific problems which were eliminated during the development efforts. Finally, some thoughts on application of more recent joining technology are presented. 78 figures

  4. Impact test of components

    Borsoi, L.; Buland, P.; Labbe, P.

    1987-01-01

    Stops with gaps are currently used to support components and piping: it is simple, low cost, efficient and permits free thermal expansion. In order to keep the nonlinear nature of stops, such design is often modeled by beam elements (for the component) and nonlinear springs (for the stops). This paper deals with the validity and the limits of these models through the comparison of computational and experimental results. The experimental results come from impact laboratory tests on a simplified mockup. (orig.)

  5. Pilot studies of management of ageing of nuclear power plant instrumentation and control components

    Burnay, S.G.; Simola, K.; Kossilov, A.; Pachner, J.

    1993-01-01

    This paper describes pilot studies which have been implemented to study the aging behavior of safety related component parts of nuclear power plants. In 1989 the IAEA initiated work on pilot studies related to the aging of such components. Four components were identified for study. They are the primary nozzle of a reactor vessel; a motor operated isolating valve; the concrete containment building; and instrumentation and control cables within the containment facility. The study was begun with phase 1 efforts directed toward understanding the aging process, and methods for monitoring and minimizing the effects of aging. Phase 2 efforts are directed toward aging studies, documentation of the ideas put forward, and research to answer questions identified in phase 1. This paper describes progress made on two of these components, namely the motor operated isolation valves, and in-containment I ampersand C cables

  6. The use of probabilistic safety assessment (PSA) based maintenance indicators to increase the availability of safety related systems in nuclear power plants

    Kirchsteiger, C.

    1991-04-01

    This work describes the theoretical development of a Probabilistic Safety Assessment (PSA) based Performance Indicator (PI) model for a comprehensive Maintenance Efficiency Analysis (MEA) and its practical application to past operational history data of a certain nuclear power plant. Plant specific equipment history and maintenance work on data have been collected and analysed using various advanced statistical procedures (nonparametric methods, multivariate analysis in order to be able to estimate safety system related equipment and maintenance process trends. The main results of such a MEA case study are the trends in the (in)effectiveness of the performance of a selected safety system and its dominant components as well as the detection of the dominant maintenance related causes of its bad (good) equipment performance. Finally, the therefrom gained results are used to propose a new set of safety system-based and maintenance-related performance indicators, including suggestions for a corresponding plant specific maintenance data collection system. (author)

  7. Technical-evaluation report on the proposed technical-specification changes for the inservice surveillance of safety-related hydraulic and mechanical snubbers at the Millstone Nuclear Power Station, Unit 2 (Docket No. 50-336)

    Selan, J.C.

    1983-01-01

    This report documents the technical evaluation of the proposed Technical Specification changes to Limiting Conditions for Operation, Surveillance Requirements and Bases for safety-related hydraulic and mechanical snubbers at the Millstone Nuclear Power Station, Unit 2. The evaluation is to determine whether the proposed Technical Specifications are in conformance with the model Standard Technical Specification set forth by the NRC. A check list, Appendix A of this report, compares the licensee's submittal with the NRC requirements and includes Proposed Resolution of the Deviations

  8. Technical-evaluation report on the proposed technical-specification changes for the inservice surveillance of safety-related hydraulic and mechanical snubbers at the Indian Point Nuclear Power Plant, Unit 3 (Docket No. 50-286)

    Selan, J.C.

    1983-01-01

    This report documents the technical evaluation of the proposed Technical Specification changes to Limiting Conditions for Operation, Surveillance Requirements and Bases for safety-related hydraulic and mechanical snubbers at the Indian Point Nuclear Power Plant, Unit 3. The evaluation is to determine whether the proposed Technical Specifications are in conformance with the model Standard Technical Specification set forth by the NRC. A check list, Appendix A of this report, compares the licensee's submittal with the NRC requirements and includes Proposed Resolution of the Deviations

  9. Technical-evaluation report on the proposed technical-specification changes for the inservice surveillance of safety-related hydraulic and mechanical snubbers at the James A. Fitzpatrick Nuclear Power Plant (Docket No. 50-333)

    Selan, J.C.

    1983-01-01

    This report documents the technical evaluation of the proposed Technical Specification changes to Limiting Conditions for Operation, Surveillance Requirements and Bases for safety-related hydraulic and mechanical snubbers at the James A. FitzPatrick Nuclear Power Plant. The evaluation is to determine whether the proposed Technical Specifications are in conformance with the model Standard Technical Specification set forth by the NRC. A check list, Appendix A of this report, compares the licensee's submittal with the NRC requirements and includes Proposed Resolution of the Deviations

  10. Summary of quantum aspects of gravitation workshop

    Summary of quantum aspects of gravitation workshop. GHANASHYAM DATE and JNANADEV MAHARANA. The Institute of Mathematical Sciences, CIT Campus, Taramani, Chennai 600 113, India. The Institute of Physics, Sachivalaya Marg, Bhubaneswar 751 005, India. Abstract. This is a summary of the presentations at ...

  11. 40 CFR 25.8 - Responsiveness summaries.

    2010-07-01

    ... required.) Responsiveness summaries shall be forwarded to the appropriate decision-making official and... 25.8 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY GENERAL PUBLIC PARTICIPATION IN... part shall prepare a Responsiveness Summary at specific decision points as specified in program...

  12. Summary Information report. Vol. 2, No. 3

    1983-08-01

    The Summary Information Report (SIR) provides summary data concerning NRC and its licensees for general use by the Chairman, other Commissioners and Commission staff offices, the Executive Director for Operations, and the Office Directors. SIR is published quarterly by the Management Information Branch (49-27834) of the Office of Resource Management

  13. 15 CFR 785.7 - Summary decision.

    2010-01-01

    ... 15 Commerce and Foreign Trade 2 2010-01-01 2010-01-01 false Summary decision. 785.7 Section 785.7 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) BUREAU OF INDUSTRY AND SECURITY, DEPARTMENT OF COMMERCE ADDITIONAL PROTOCOL REGULATIONS ENFORCEMENT § 785.7 Summary...

  14. Detecting Terrorism Incidence Type from News Summary

    Nizamani, Sarwat; Memon, Nasrullah

    2012-01-01

    The paper presents the experiments to detect terrorism incidence type from news summary data. We have applied classification techniques on news summary data to analyze the incidence and detect the type of incidence. A number of experiments are conducted using various classification algorithms...... and results show that a simple decision tree classifier can learn incidence type with satisfactory results from news data....

  15. Pilot Research Summaries, 1967-1970.

    Casey, James L.; Hayes, Larry K.

    This report contains one-page summaries of a majority of the 134 research studies funded through the Oklahoma Consortium on Research Development. The research covers the whole spectrum of academic topics , from nursing to ecology to art to politics.. Brief summaries of a majority of the 37 development seminars funded through the Consortium are…

  16. 15 CFR 904.505 - Summary sale.

    2010-01-01

    ... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Summary sale. 904.505 Section 904.505 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) NATIONAL OCEANIC... and Forfeiture Procedures § 904.505 Summary sale. (a) In view of the perishable nature of fish, any...

  17. 40 CFR 35.532 - Requirements summary.

    2010-07-01

    ... 40 Protection of Environment 1 2010-07-01 2010-07-01 false Requirements summary. 35.532 Section 35.532 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY GRANTS AND OTHER FEDERAL ASSISTANCE....532 Requirements summary. (a) Applicants and recipients of Performance Partnership Grants must meet...

  18. 40 CFR 35.132 - Requirements summary.

    2010-07-01

    ... 40 Protection of Environment 1 2010-07-01 2010-07-01 false Requirements summary. 35.132 Section 35.132 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY GRANTS AND OTHER FEDERAL ASSISTANCE... Requirements summary. Applicants and recipients of Performance Partnership Grants must meet: (a) The...

  19. Summary Report for Online Schools and Programs

    Colorado Department of Education, 2014

    2014-01-01

    Pursuant to State Law, the Colorado Department of Education, Office of Blended and Online Learning is required to prepare an annual summary report for submission. The passage of a later State House Bill repealed the annual requirement for the Summary Report and also the annual reporting mandates that were required of all online schools and…

  20. 12 CFR 1806.101 - Summary.

    2010-01-01

    ... 12 Banks and Banking 7 2010-01-01 2010-01-01 false Summary. 1806.101 Section 1806.101 Banks and Banking COMMUNITY DEVELOPMENT FINANCIAL INSTITUTIONS FUND, DEPARTMENT OF THE TREASURY BANK ENTERPRISE AWARD PROGRAM General Provisions § 1806.101 Summary. (a) Under the Bank Enterprise Award Program, the...